SEC Petition Evaluation Report 

Petition SEC-00126 

Report Rev #:_ 0__ Report Submittal Date: April 27, 2009 

Subject Expert(s): Louise Buker, Roger Halsey, Karin Jessen, Dan Mantooth 

Site Expert(s): N/A 

Petition Administrative Summary 

Petition Under Evaluation 

Petition # Petition 

Type 

Petition 

Qualification Date 

DOE/AWE Facility Name 

SEC-00126 83.13 December 22, 2008 Piqua Organic Moderated Reactor (POMR) 

Petitioner Class Definition 

All employees associated with reactor activities who worked within and around the reactor dome at the Piqua 

Organic Moderated Reactor during the covered period from January 1, 1963 through December 31, 1966. 

Class Evaluated by NIOSH 

All employees associated with reactor activities who worked within and around the reactor dome at the Piqua 

Organic Moderated Reactor during the covered period from January 1, 1963 through May 1, 1966. 

NIOSH-Proposed Class to be Added to the SEC 

None 

Related Petition Summary Information 

SEC Petition Tracking #(s) Petition Type DOE/AWE Facility Name Petition Status 

N/A N/A N/A N/A 

Related Evaluation Report Information 

Report Title DOE/AWE Facility Name 

N/A N/A 

ORAU Lead Technical Evaluator: Karin Jessen ORAU Review Completed By: Daniel Stempfley 

Peer Review Completed By: [Signature on file] 4/27/2009 

Charles D. Nelson Date 

SEC Petition Evaluation Reviewed By: [Signature on file] 4/27/2009 

J. W. Neton Date 

SEC Evaluation Approved By: [Signature on file] 4/28/2009 

Dave Sundin for Larry Elliott Date 
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Evaluation Report Summary: SEC-00126, Piqua Organic Moderated Reactor 

This evaluation report by the National Institute for Occupational Safety and Health (NIOSH) 

addresses a class of employees proposed for addition to the Special Exposure Cohort (SEC) per the 

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended, 42 U.S.C. 

§ 7384 et seq. (EEOICPA) and 42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. 

Petitioner-Requested Class Definition

Petition SEC-00126, qualified on December 22, 2008, requested that NIOSH consider the following 

class: All employees associated with reactor activities who worked within and around the reactor 

dome at the Piqua Organic Moderated Reactor during the covered period from January 1, 1963 

through December 31, 1966.

Class Evaluated by NIOSH

Based on its preliminary research, NIOSH reduced the petitioner-requested class. NIOSH evaluated 

the following class: All employees associated with reactor activities who worked within and around 

the reactor dome at the Piqua Organic Moderated Reactor during the covered period from January 1, 

1963 through May 1, 1966. The petitioner-requested class was modified because the Piqua operations 

that began on May 2, 1966 (the start of the Recovery Program) were significantly different, perhaps 

with more potential exposures, than those occurring during the operational period from January 1, 

1963 through May 1, 1966. NIOSH has identified information regarding the Piqua Recovery Program 

that warrants further research. During the Recovery Program the reactor facility was shutdown and a 

portion of the reactor and associated shielding were removed to inspect the reactor core and facilitate 

the decontamination of the coke-like mass located in the core of reactor. This activity was not 

associated with normal operational activities that began in 1963 and continued into 1966. Therefore, 

the review of the May 2, 1966 through December 31, 1966 timeframe has been reserved in this 

evaluation report, pending further review and research. In addition, NIOSH is awaiting a Department 

of Labor (DOL) decision regarding extending the covered period (currently defined by the DOE 

Office of Health, Safety and Security as starting in 1963 and ending in 1966) so that it can determine 

if the Decontamination and Decommissioning activities that were completed in February 1969 are 

covered activities.

NIOSH-Proposed Class to be Added to the SEC

Based on its full research of the class under evaluation, NIOSH has obtained monthly, semiannual, 

and annual reports, as well as information from former Piqua Organic Moderated Reactor (POMR) 

facility personnel that provide sufficient information to allow dose reconstructions to be performed 

with sufficient accuracy. Based on its analysis of these available resources, NIOSH found no part of 

the class under evaluation for which it cannot estimate radiation doses with sufficient accuracy. 
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Feasibility of Dose Reconstruction

Per EEOICPA and 42 C.F.R. § 83.13(c)(1), NIOSH has established that it has access to sufficient 

information to: (1) estimate the maximum radiation dose, for every type of cancer for which radiation 

doses are reconstructed, that could have been incurred in plausible circumstances by any member of 

the class; or (2) estimate radiation doses of members of the class more precisely than an estimate of 

maximum dose. Information available is sufficient to document or estimate the maximum internal 

and external potential exposure to members of the evaluated class under plausible circumstances 

during the specified period. 

Health Endangerment Determination

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is not required 

because NIOSH has determined that it has sufficient information to estimate dose for the members of 

the evaluated class. 
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Table of Contents 

Evaluation Report Summary: SEC-00126, Piqua Organic Moderated Reactor...................................... 3

1.0 Purpose and Scope.......................................................................................................................... 7

2.0 Introduction .................................................................................................................................... 7

3.0 SEC-00126, Piqua Organic Moderated Reactor (POMR) Class Definitions ................................. 8 

3.1 Petitioner-Requested Class Definition and Basis.................................................................. 8 

3.2 Class Evaluated by NIOSH ................................................................................................... 9 

3.3 NIOSH-Proposed Class to be Added to the SEC ................................................................ 10

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class........................................................... 10 

4.1 ORAU Technical Information Bulletin (OTIB) .................................................................. 10 

4.2 Facility Employees and Experts .......................................................................................... 10 

4.3 Previous Dose Reconstructions ........................................................................................... 12 

4.4 NIOSH Site Research Database .......................................................................................... 12 

4.5 Documentation and/or Affidavits Provided by Petitioners ................................................. 12

5.0 Radiological Operations Relevant to the Class Evaluated by NIOSH ......................................... 13 

5.1 POMR Plant and Process Descriptions ............................................................................... 13 

5.2 Radiological Exposure Sources from POMR Operations ................................................... 21 

5.2.1 Internal Radiological Exposure Sources from POMR Operations.......................... 21 

5.2.1.1 Activated Impurities in the Coolant ......................................................... 22 

5.2.1.2 Activated Corrosion Products................................................................... 22 

5.2.1.3 Neutron Recoil Reaction Products ........................................................... 23 

5.2.1.4 Fission Products........................................................................................ 23 

5.2.2 External Radiological Exposure Sources from POMR Operations......................... 24 

5.2.2.1 Photon....................................................................................................... 25 

5.2.2.2 Beta........................................................................................................... 25 

5.2.2.3 Neutron..................................................................................................... 25

6.0 Summary of Available Monitoring Data for the Class Evaluated by NIOSH ............................. 26 

6.1 Available POMR Internal Monitoring Data ........................................................................ 26 

6.2 Available POMR External Monitoring Data....................................................................... 27

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by NIOSH ...................................... 31 

7.1 Pedigree of POMR Data...................................................................................................... 32 

7.1.1 Internal Monitoring Data Pedigree Review............................................................. 32 

7.1.2 External Monitoring Data Pedigree Review............................................................ 33 

7.2 Evaluation of Bounding Internal Radiation Doses at POMR.............................................. 33 

7.2.1 Evaluation of Bounding Process-Related Internal Doses........................................ 34 

7.2.2 Evaluation of Bounding Ambient Environmental Internal Doses........................... 35 

7.2.3 Methods for Bounding Internal Dose at POMR...................................................... 35 

7.2.3.1 Methods for Bounding Operational Period Internal Dose........................ 35 
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7.2.3.2 Methods for Bounding Ambient Environmental Internal Dose ............... 37 

7.2.4 Internal Dose Reconstruction Feasibility Conclusion ............................................. 37 

7.3 Evaluation of Bounding External Radiation Doses at POMR............................................. 38 

7.3.1 Evaluation of Bounding Process-Related External Doses....................................... 38 

7.3.1.1 Personnel Dosimetry Data........................................................................ 38 

7.3.1.2 Area Monitoring Data............................................................................... 40 

7.3.2 Evaluation of Bounding Ambient Environmental External Doses.......................... 41 

7.3.3 POMR Occupational X-Ray Examinations............................................................. 41 

7.3.4 Methods for Bounding External Dose at POMR..................................................... 42 

7.3.4.1 Methods for Bounding Operational Period External Dose....................... 42 

7.3.5 External Dose Reconstruction Feasibility Conclusion............................................ 43 

7.4 Evaluation of Petition Basis for SEC-00126....................................................................... 44 

7.4.1 Covered Period of Employment .............................................................................. 44 

7.4.2 Monitoring............................................................................................................... 44 

7.5 Summary of Feasibility Findings for Petition SEC-00126.................................................. 44

8.0 Evaluation of Health Endangerment for Petition SEC-00126...................................................... 45

9.0 Class Conclusion for Petition SEC-00126 ................................................................................... 45

10.0 References .................................................................................................................................... 47

Attachment One: Information from Interviews with Former Piqua Personnel ..................................... 51 

Attachment Two: Data Capture Synopsis ............................................................................................. 55 

Tables 

Table 4-1: No. of POMR Claims Submitted Under the Dose Reconstruction Rule ............................ 12

Table 5-1: Summary of POMR Operating History ............................................................................... 14 

Table 5-2: Radionuclides of Concern for Internal Exposure................................................................. 24

Table 6-1: Piqua Summary Annual Whole-Body Penetrating Exposures............................................. 27 

Table 6-2: Piqua Semiannual and Monthly Report Results .................................................................. 28

Table 7-1: Radionuclides of Concern by Category ............................................................................... 36 

Table 7-2: Summary of Feasibility Findings for SEC-00126 ............................................................... 45 

Figure 

Figure 5-1: Piqua Core I Fuel Element.................................................................................................. 16 
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SEC Petition Evaluation Report for SEC-00126

1.0 Purpose and Scope 

This report evaluates the feasibility of reconstructing doses for all employees associated with reactor 

activities who worked within and around the reactor dome at the Piqua Organic Moderated Reactor 

during the covered period from January 1, 1963 through May 1, 1966. It provides information and 

analyses germane to considering a petition for adding a class of employees to the congressionallycreated SEC. 

This report does not make any determinations concerning the feasibility of dose reconstruction that 

necessarily apply to any individual energy employee who might require a dose reconstruction from 

NIOSH. This report also does not contain the final determination as to whether the proposed class 

will be added to the SEC (see Section 2.0). 

This evaluation was conducted in accordance with the requirements of EEOICPA, 42 C.F.R. pt. 83, 

and the guidance contained in the Office of Compensation Analysis and Support’s (OCAS) Internal 

Procedures for the Evaluation of Special Exposure Cohort Petitions, OCAS-PR-004. 

2.0 Introduction 

Both EEOICPA and 42 C.F.R. pt. 83 require NIOSH to evaluate qualified petitions requesting that the 

Department of Health and Human Services (HHS) add a class of employees to the SEC. The 

evaluation is intended to provide a fair, science-based determination of whether it is feasible to 

estimate with sufficient accuracy the radiation doses of the class of employees through NIOSH dose 

reconstructions.1

 

42 C.F.R. § 83.13(c)(1) states: Radiation doses can be estimated with sufficient accuracy if NIOSH 

has established that it has access to sufficient information to estimate the maximum radiation dose, 

for every type of cancer for which radiation doses are reconstructed, that could have been incurred in 

plausible circumstances by any member of the class, or if NIOSH has established that it has access to 

sufficient information to estimate the radiation doses of members of the class more precisely than an 

estimate of the maximum radiation dose.

 

Under 42 C.F.R. § 83.13(c)(3), if it is not feasible to estimate with sufficient accuracy radiation doses 

for members of the class, then NIOSH must determine that there is a reasonable likelihood that such 

radiation doses may have endangered the health of members of the class The regulation requires 

NIOSH to assume that any duration of unprotected exposure may have endangered the health of 

members of a class when it has been established that the class may have been exposed to radiation 

during a discrete incident likely to have involved levels of exposure similarly high to those occurring 

during nuclear criticality incidents. If the occurrence of such an exceptionally high-level exposure has 

not been established, then NIOSH is required to specify that health was endangered for those workers 

 

1

 NIOSH dose reconstructions under EEOICPA are performed using the methods promulgated under 42 C.F.R. pt. 82 and 

the detailed implementation guidelines available at http://www.cdc.gov/niosh/ocas. 
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who were employed for at least 250 aggregated work days within the parameters established for the 

class or in combination with work days within the parameters established for other SEC classes 

(excluding aggregate work day requirements). 

NIOSH is required to document its evaluation in a report, and to do so, relies upon both its own dose 

reconstruction expertise as well as technical support from its contractor, Oak Ridge Associated 

Universities (ORAU). Once completed, NIOSH provides the report to both the petitioner(s) and to the 

Advisory Board on Radiation and Worker Health (Board). The Board will consider the NIOSH 

evaluation report, together with the petition, petitioner(s) comments, and other information the Board 

considers appropriate, in order to make recommendations to the Secretary of HHS on whether or not 

to add one or more classes of employees to the SEC. Once NIOSH has received and considered the 

advice of the Board, the Director of NIOSH will propose a decision on behalf of HHS. The Secretary 

of HHS will make the final decision, taking into account the NIOSH evaluation, the advice of the 

Board, and the proposed decision issued by NIOSH. As part of this decision process, petitioners may 

seek a review of certain types of final decisions issued by the Secretary of HHS.2

 

3.0 SEC-00126, Piqua Organic Moderated Reactor (POMR) Class 

Definitions 

The following subsections address the evolution of the class definition for SEC-00126, the Piqua 

Organic Moderated Reactor (POMR). When a petition is submitted, the requested class definition is 

reviewed as submitted. Based on its review of the available site information and data, NIOSH will 

make a determination whether to qualify for full evaluation all, some, or no part of the petitionerproposed class. If some portion of the petitioner-proposed class is qualified, NIOSH will specify that 

class along with a justification for any modification of petitioner’s class. After a full evaluation of the 

qualified class, NIOSH will determine whether to propose a class for addition to the SEC and will 

specify that proposed class definition. 

3.1 Petitioner-Requested Class Definition and Basis 

Petition SEC-00126, qualified on December 22, 2008, requested that NIOSH consider the following 

class for addition to the SEC: All employees associated with reactor activities who worked within and 

around the reactor dome at the Piqua Organic Moderated Reactor during the covered period from 

January 1, 1963 through December 31, 1966. 

The petitioner provided information and affidavit statements in support of the petitioner’s belief that 

accurate dose reconstruction over time is impossible for the POMR workers in question. NIOSH 

deemed the following information and affidavit statements sufficient to qualify SEC-00126 for 

evaluation: 

In support of the petition, the SEC-00126 petitioner claims that no records were kept on 

activities related to the Piqua site to include the dismantling of the facility and that the 

 

2

 See 42 C.F.R. pt. 83 for a full description of the procedures summarized here. Additional internal procedures are 

available at http://www.cdc.gov/niosh/ocas. 
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petitioner’s father was not trained in the hazards of the POMR. In addition, no monitoring 

devices were ever offered.

During the construction and operation of the reactor, [former POMR employee] was 

employed by the city of Piqua as a material handler and subsequently a linesman for the 

power plant and helped out at the reactor site at various times even though he was primarily 

assigned to the original power plant adjacent to one another. He was a laborer primarily 

assigned any duty his supervisor would assign. This all done without any type of monitoring, 

training, or protective devices for handling nuclear material [sic] this was not explained but 

one has to surmise it was because of [former POMR employee]’s primary work was at the 

other building [sic]. It was during the second part of his career, decommissioning and 

dismantling of the Reactor, that [former POMR employee] had the most potential exposure to 

the harmful elements used at and remain at the Moderated Reactor located in Piqua Ohio. 

... At no time was [name of the former Energy employee] trained in the hazards [at the POMR 

site] nor was any monitoring devices offered.” “No records were kept on activities related to 

the Piqua site to include the dismantling of the facility and use of [former POMR employee 

and survivor’s father] in such exposure. 

Based on its POMR research and data capture efforts, NIOSH determined that it has access to 

monthly, semiannual, and annual summary reports, reactor design, shielding material, and radiation 

source information for the time period under evaluation, but NIOSH has also determined that internal 

and external records are not complete for all time periods or for all radionuclides. The information 

and statements provided by the petitioner qualified the petition for further consideration by NIOSH, 

the Board, and HHS. The details of the petition basis are addressed in Section 7.4. 

 

3.2 Class Evaluated by NIOSH 

Based on its preliminary research, NIOSH reduced the petitioner-proposed class to include only the 

January 1, 1963 through May 1, 1966 timeframe, because the Piqua operations that began on May 2, 

1966 (the start of the Recovery Program) were significantly different, perhaps with more potential 

exposures than those that occurred during the operational period from January 1, 1963 through May 1, 

1966. NIOSH has identified information regarding the Piqua Recovery Program that warrants further 

research. During the Recovery Program, the reactor was shutdown and a portion of the reactor and 

associated shielding were removed to inspect the reactor core and facilitate the decontamination of the 

coke-like mass located in the core of reactor. This activity was not associated with normal operational 

activities that began in 1963 and continued through May 1, 1966. Therefore, the review of the May 2, 

1966 through December 31, 1966 timeframe has been reserved in this evaluation report, pending 

further review and research. In addition, NIOSH is awaiting a Department of Labor (DOL) decision 

regarding extending the covered period (currently defined by the DOE Office of Health, Safety and 

Security as starting in 1963 and ending in 1966) so that it can determine if the Decontamination and 

Decommissioning activities that were completed in February 1969 are covered activities. NIOSH 

defined the following class for further evaluation: All employees associated with reactor activities 

who worked within and around the reactor dome at the Piqua Organic Moderated Reactor during the 

covered period from January 1, 1963 through May 1, 1966.
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3.3 NIOSH-Proposed Class to be Added to the SEC 

Based on its research, NIOSH has obtained monthly, semiannual, and annual reports, limited radiation 

exposure records, and information from former POMR facility personnel that allow dose 

reconstruction to be performed with sufficient accuracy. Based on its analysis of these available 

resources, NIOSH found no part of the class under evaluation for which it cannot estimate radiation 

doses with sufficient accuracy. 

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class 

In addition to searching City of Piqua records, NIOSH completed an extensive database and Internet 

search for information regarding the Piqua Organic Moderated Reactor facility. The database search 

included the DOE Legacy Management Considered Sites database, the DOE Office of Scientific and 

Technical Information (OSTI) database, the Energy Citations database, the Atomic Energy Technical 

Report database, and the Hanford Declassified Document Retrieval System. In addition to general 

Internet searches, the NIOSH Internet search included OSTI OpenNet Advanced searches, OSTI 

Information Bridge Fielded searches, Nuclear Regulatory Commission (NRC) Agency-wide 

Documents Access and Management (ADAMS) web searches, the DOE Office of Human Radiation 

Experiments website, and the DOE-National Nuclear Security Administration-Nevada Site Officesearch. Attachment Two contains a summary of Piqua Organic Moderated Reactor documents. The 

summary specifically identifies data capture details and general descriptions of the documents 

retrieved. 

In addition to the database and Internet searches listed above, NIOSH identified and reviewed 

numerous data sources to determine information relevant to determining the feasibility of dose 

reconstruction for the class of employees under evaluation. This included determining the availability 

of information on personal monitoring, area monitoring, industrial processes, and radiation source 

materials. The following subsections summarize the data sources identified and reviewed by NIOSH. 

4.1 ORAU Technical Information Bulletin (OTIB) 

An ORAU Technical Information Bulletin (OTIB) is a general working document that provides 

guidance for preparing dose reconstructions at particular sites or categories of sites. NIOSH reviewed 

the following OTIB as part of its evaluation: 

• OTIB: Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures, ORAUTOTIB-0006; December 21, 2005; SRDB Ref ID: 20220

4.2 Facility Employees and Experts 

To obtain additional information, NIOSH contacted seven former POMR facility personnel. One 

person contacted did not work at the POMR facility during the time period evaluated in this report; 

therefore, NIOSH did not conduct a complete interview with this person. NIOSH performed seven 

interviews with six former Piqua employees (one former employee was interviewed on two separate 
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occasions), all were considered to be knowledgeable about the POMR facility. All interviews were 

conducted by phone. The purpose of the interviews was to gain additional first-hand information 

from people who worked at the POMR facility. A summary of the information obtained from the 

interviews can be found in Attachment One of this report. 

• Personal Communication, 2009a, Personal Communication with Former Health Physics 

Technician; Telephone Interview by ORAU Team; January 20, 2009; SRDB Ref ID: 61683

• Personal Communication, 2009b, Personal Communication with DOE Legacy Management 

Employee; Telephone Interview by ORAU Team; January 29, 2009; SRDB Ref ID: 61681 

• Personal Communication, 2009c, Personal Communication with Former Shift Supervisor; 

Telephone Interview by ORAU Team; February 19, 2009; SRDB Ref ID: 61677

• Personal Communication, 2009d, Personal Communication with Former Construction Engineer, 

Instrumentation Engineer, and Electrical Engineer; Telephone Interview by ORAU Team; 

February 19, 2009; SRDB Ref ID: 61684 

• Personal Communication, 2009e, Personal Communication with Former Reactor Operator and 

Maintenance Foreman; Telephone Interview by ORAU Team; February 6, 2009; SRDB Ref ID: 

61679

• Personal Communication, 2009f, Personal Communication with Chief Health Physicist; 

Telephone Interview by ORAU Team; February 23, 2009; SRDB Ref ID: 61680

• Personal Communication, 2009g, Personal Communication with Former Health Physics 

Technician, second interview; Telephone Interview by ORAU Team; March 18, 2009; SRDB Ref 

ID: 62597

• Personal Communication, 2009h, Personal Communication with Former POMR Facility 

Employee; documented telephone communication; March 23, 2009; SRDB Ref ID: 62596
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4.3 Previous Dose Reconstructions 

NIOSH reviewed its NIOSH OCAS Claims Tracking System (NOCTS) to locate EEOICPA-related 

dose reconstructions that might provide information relevant to the petition evaluation. Table 4-1 

summarizes the results of this review. (NOCTS data available as of April 21, 2009) 

Table 4-1: No. of POMR Claims Submitted Under the Dose Reconstruction Rule 

Description Totals 

Total number of claims submitted for dose reconstruction 5 

Total number of claims submitted for energy employees who meet the definition criteria for the class 

under evaluation (January 1, 1963 through May 1, 1966) 5 

Number of dose reconstructions completed for energy employees who meet the definition criteria for 

the class under evaluation (i.e., the number of such claims completed by NIOSH and submitted to the 

Department of Labor for final approval). 3 

Number of claims for which internal dosimetry records were obtained for the identified years in the 

evaluated class definition 0 

Number of claims for which external dosimetry records were obtained for the identified years in the 

evaluated class definition 1 

NIOSH reviewed each claim to determine whether internal and/or external personal monitoring 

records could be obtained for the employee. Five claims have been submitted for the POMR facility; 

three of the claims have been completed. NIOSH has located no claims with internal monitoring and 

only one claim with external monitoring for individuals that were employed during the period 

evaluated in this report. 

4.4 NIOSH Site Research Database 

NIOSH also examined its Site Research Database (SRDB) to locate documents supporting the 

evaluation of the proposed class. One hundred ninety-five documents in this database were identified 

as pertaining to the Piqua Organic Moderated Reactor facility. These documents were evaluated for 

their relevance to this petition. The documents include personnel data, historical background 

information on the Piqua facility, and monthly, semiannual, and annual reports. Through researching 

the records, NIOSH learned that R.S. Landauer Jr. and Company provided external dosimetry. 

Although R.S. Landauer Jr. and Company has stated that it cannot locate any POMR records, NIOSH 

is continuing its efforts to work with R.S. Landauer Jr. and Company regarding potential POMR data. 

4.5 Documentation and/or Affidavits Provided by Petitioners 

In qualifying and evaluating the petition, NIOSH reviewed the following document submitted by the 

petitioner: 

• Petition Form B; August 21, 2008; OSA Ref ID: 106811 
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5.0 Radiological Operations Relevant to the Class Evaluated by 

NIOSH 

The following subsections summarize both radiological operations at the POMR site from January 1, 

1963 through May 1, 1966 and the information available to NIOSH to characterize particular 

processes and radioactive source materials. From available sources NIOSH has gathered monthly, 

semiannual, and annual summary reports, reactor design information, and shielding and radiation 

source materials. The information included within this evaluation report is intended only to be a 

summary of the available information. As previously discussed, NIOSH’s review and evaluation of 

the May 2, 1966 through December 31, 1966 timeframe is reserved. Therefore, only the radiological 

operations performed January 1, 1963 through May 1, 1966 will be reviewed throughout this report. 

5.1 POMR Plant and Process Descriptions 

ATTRIBUTION: Section 5.1 was completed by Karin Jessen, Oak Ridge Associated Universities 

(ORAU). These conclusions were peer-reviewed by the individuals listed on the cover page. The 

rationales for all conclusions in this document are explained in the associated text.

The POMR site, also known as the Piqua Nuclear Power facility, was a nuclear power plant designed 

by Atomics International, that operated in Piqua, Ohio, about 34 miles north of Dayton, Ohio. The 

plant contained a 45.5-megawatt (thermal) organically cooled and moderated nuclear reactor that was 

originally built as a demonstration project by the AEC. The reactor was initially operated by Atomics 

International, who also trained employees from the City of Piqua to operate and maintain the plant. 

The City of Piqua eventually took over operation of the reactor, which operated between 1963 and 

1966. Operations were terminated in 1966, due to technical difficulties such as control-rod problems 

and fouling of heat-transfer surfaces; in 1967, the AEC terminated its contract for reactor operation 

with the City of Piqua. The Piqua reactor was dismantled between 1967 and 1969, and the radioactive 

coolant and most other radioactive materials were removed. The remaining radioactive structural 

components of the reactor were entombed in the reactor vessel under sand and concrete (DOE, 2009). 

The primary objective of the POMR facility was to demonstrate the feasibility of the organic reactor 

concept when operated as an integral part of a power generation system, and to provide information 

which ultimately was expected to lead to the design and construction of larger, more economically 

competitive nuclear power facilities (DOE, 2009). The plant was designed as a load-following 

system, meaning the reactor power level varied according to the steam demands. Superheated steam 

at constant pressure was supplied to the steam header in the Piqua Municipal Power Plant, located to 

the west across the Great Miami River. A summary of the POMR operating history is shown in Table 

5-1. 
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Table 5-1: Summary of POMR Operating History

Date Action 

June 1963 Initially criticality achieved. 

July 1963 Fuel loading completed. 

January 27, 1964 Full power achieved; reactor operated steadily but with one scram. 

May 21, 1964 First scheduled shutdown for routine maintenance and inspection 

During this period of operation, POMR contributed ~ 40% of the energy generated by the 

City of Piqua. 

December 7, 1964 Reactor was shut down to renew fifteen in-vessel filters and remove the fuel element in 

Core position F-13 for examination. 

January 28, 1965 Reactor was shut down for complete replacement of in-vessel filters, maintenance, and for 

relocation of the instrumented fuel element from position E-12 to position D-5. 

April 2, 1965 Several malfunctioning control rod drive units repaired. 

Concern over possible plugged condition of the inner process tube of the control rodbearing elements led to the movement of the six inner ring control rod elements to 

peripheral positions. The core size was increased from 61- 67 fuel elements. 

May 6 – 12, 1965 Scram occurred on May 6, 1965. During this time, the reactor coolant level had been 

lowered by operational error, which resulted in a temporary loss of circulation through 

three elements. 

Shutdown was extended until May 12 so the three fuel elements could be removed to 

spent-fuel storage. 

May 13, 1965 (estimated 

date) 

Immediately upon restart, excessive surface temperatures were noted, necessitating 

additional fuel element removal. 

Because of the fuel element removal, the system operated with only one coolant pump 

during the latter half of June and into July. 

July 18, 1965 Reactor shut down for modifications, maintenance, and in-vessel filter replacement; 

performed extensive modifications of the in-core control rod circuitry. 

September 6, 1965 Reactor operation resumed. 

October 12, 1965 Reactor shut down, fuel rearrangements were made, increasing the core loading to 70 fuel 

elements. 

October 23, 1965 Reactor restarted. Operation of the reactor continued at an average power level of about 

24 MWt. 

January 13, 1966 Reactor scrammed because of a spurious signal. At this time, there was no indication of 

any unusual condition in the reactor core. Prior to restarting the reactor, an abnormal incore condition was identified during the performance of a rod-drop test. 

Note: The reactor was shut down sometime after the abnormal in-core condition.

Notes: 

This document was created from information included in Summary of Operating History (Unknown author, unknown datea) and from information included in Piqua Recovery Program—Phase II-A Plan of Action (Morgan, 1966). 

Three pipelines, approximately 1,400 feet in length, connected the POMR facility on the east side of 

the Great Miami River and the Municipal Power Plant on the west side of the river. These three lines 

included: (1) a 12-inch diameter steam line for steam flow from the POMR steam generator to the 

conventional power plant steam header. This line also provided preheat and start-up steam from the 

conventional plant to the POMR facility, (2) a 6-inch diameter line for boiler feed-water supply from 

the conventional plant to the POMR facility, and (3) a 3-inch line for the return of process steam 

condensate from the POMR facility to the conventional plant. 
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POMR Components 

Reactor Vessel 

The reactor vessel was a pressure vessel made of low-carbon steel (SA-212-B) and had an internal 

diameter of 7.6 feet, an overall height of 27 feet, and an average wall thickness of 2 inches. It was 

designed to meet applicable American Society of Mechanical Engineers’ codes for an internal 

pressure of 300 psia and 750°F. The vessel contained ten coolant inlet and outlet nozzle penetrations 

that were welded and flanged to the vessel. The penetrations consisted of a single 20-inch coolant 

inlet nozzle, two 14-inch coolant outlet nozzles, a single 6-inch auxiliary inlet nozzle, a 6-inch 

auxiliary outlet nozzle, two 14-inch sampling valve nozzles, two 8-inch control rod cable nozzles, and 

one 8-inch thermocouple lead-through nozzle (Atomics International, 1965a, p. 5). 

Reactor Core 

The core was positioned near the bottom of the reactor core tank. It consisted of a maximum of 85 

fuel elements (the number of fuel elements varied during operations) of slightly enriched uranium and 

13 control-safety rods and was surrounded by an annular thermal shield supported from a ledge inside 

the tank. Steel grid plates, located above and below the core, supported the fuel elements and control 

rods. 

The core pressure was maintained by a control valve in the line to the degasifier. Regulated 

pressurizing pumps returned a constant flow from the degasifier into the primary loop. One 

pressurizing pump had to be running at all times; the second pump could be removed for service. In 

the event of loss of power to the main coolant pump, the pressurizing pumps maintained circulation 

through the core (Atomics International, 1965a, p. 8). 

Shielding

Reactor radial shielding was achieved through the following: (1) inner thermal shield with 1.5 inch 

steel; (2) outer thermal shield with 4.0 inch steel; (3) reactor vessel wall with 1.125 inch steel; and (4) 

biological shield with 8 feet 4 inches of ordinary concrete (Atomics International, 1965a). 

Shielding at the bottom of the reactor was achieved through the following: (1) the lower grid plate 

with 6-inch steel; (2) organic coolant, measuring 2.5 feet; (3) reactor vessel lower head with 1.125 

inch steel; and (4) vessel support with 3 feet of concrete resting on the reactor building lower-floor 

level (Atomics International, 1965a). 

Shielding above the reactor was achieved through the following: (1) the upper grid plate with 8 inch 

steel; (2) organic coolant, measuring 17.5 feet; and (3) reactor vessel head with 8.5 inch steel 

(Atomics International, 1965a). 

Shielding was provided in other areas that contained process fluids; these areas included the 

purification room, the drain tank room, the waste fired boiler room, and the decay tank room (Atomics 

International, 1965a). 

Control Rods 

The control rod drives consisted of compact unitized assemblies located inside the core tank and 

immersed in the coolant above the core. The neutron absorber element consisted of an assembly of 
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outside-diameter tubes filled with boron carbide, positioned to operate inside 13 selected circular fuel 

elements. The drive mechanism operated on the magnetic jack principle, wherein the rod is raised or 

lowered in discrete steps by energizing appropriate sets of electromagnetic coils. 

Fuel Elements 

The fuel elements were circular in cross-section with an outside diameter of 5.25 inches and a length 

of approximately 80 inches (Unknown author, unknown date-b). The uranium inside the elements 

was in the form of two concentric tubes. To improve heat transfer, the inner and outer surfaces of 

each fuel tube were covered with finned aluminum cladding. To maintain physical separation of the 

fuel cylinders, the fins of the cladding were twisted into a slight spiral shape along their longitudinal 

axis. The finned fuel tubes were enclosed between two concentric stainless steel tubes. The ends of 

the steel tubes were fastened to upper and lower end pieces of the fuel elements. The upper end piece 

fit into the upper grid plate and supported the weight of the fuel element. The lower end piece guided 

the element in the lower grid plate and aligned it within the core. 

Fuel elements consisted of uranium fuel, clad with aluminum. The fuel material was a metallic 

uranium alloy, enriched to approximately 1.9 weight percent uranium-235. The alloy was composed 

of uranium, approximately 3.5 weight-percent molybdenum and 0.1 weight-percent aluminum. The 

aluminum cladding had a finned surface to provide an extended heat transfer area. It was 

metallurgically bonded to the uranium fuel using a diffusion barrier of nickel approximately 0.001 

inches thick. The maximum total mass of fuel in the core was 134 kg of uranium-235 and 6,776 kg of 

uranium-238. Figure 5-1 shows a Piqua fuel element. 

An orifice was located in the inlet (upper) end of each fuel element that did not have a control rod 

associated with it. The orifices were adjustable during shutdown and were set to equalize the 

temperature rise across each fuel element. 

Figure 5-1: Piqua Core I Fuel Element 

Source: Irradiation Performance of Piqua Core 1 Fuel Elements (Arnold, 1966, p. 13) 
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Reactor Organic Coolant 

The organic coolant used in the POMR was a commercially available hydrocarbon mixture of the 

three isomers of terphenyl3. After extensive testing of many hydrocarbons, this hydrocarbon mixture 

was selected as the moderator-coolant because terphenyls exhibit relatively high thermal and radiation 

stability, are noncorrosive, and have a relatively low vapor pressure at their expected operating 

temperatures (Unknown author, unknown date-b). 

The organic coolant entered the tank above the core and flowed downward through the fuel elements 

into the lower plenum, below the lower grid plate. The coolant then flowed upward through the 

annulus between the core tank and the thermal shield into the outlet plenum of the core tank and back 

to the primary coolant loop. The organic coolant filled all of the available space in the tank and 

served as a moderator, coolant, reflector, and shield for the core. 

During full-power operation, the coolant was heated in the core from 519°F to 575°F while 

transferring 155 x 106 Btu/hr from the core. The coolant was pumped by two main coolant pumps to 

the super heater and steam boiler where the heat was transferred to the steam system. A total of 

150,000 lb/hr of superheated steam was produced at a pressure of 450 psia and a temperature of 

550°F. The main coolant system consisted of a single loop in which two 6,000 gpm pumps operated 

in parallel, pumping 12,000 gpm to a single super heater and boiler. A flow bypass was utilized to 

divert coolant around the boiler for control purposes. 

Radiation damage to the coolant, as well as exposure to heat, resulted in the formation of various 

gases and low molecular weight compounds, and at the same time produced some higher molecular 

weight compounds (referred to as high boilers). The higher molecular weight compounds consisted 

primarily of long-chained polymerization products and were removed from the coolant by vacuum 

distillation. The high boiler content was permitted to build up to about 30% in the coolant, which was 

thought to be the optimum concentration. The optimization was based on a balance of the decrease in 

radiation and thermal damage (decomposition) rate, and the decrease in the heat transfer 

characteristics of the coolant, as the high boiler content was increased (Unknown author, unknown 

date-b). 

Reactor Systems 

There were a number of auxiliary or supporting systems associated with the operation of the POMR 

facility. Three examples of auxiliary functions included (1) pressurization of the main coolant loop, 

including the reactor vessel, (2) dissipation of the residual or "decay" heat from the reactor core in the 

event the main coolant loop was inoperative, and (3) removal of water, gases, and other low-boiling 

materials formed in the core as a result of coolant decomposition. The schematics associated with the 

POMR systems are relayed in Final Safeguards Summary Report for the Piqua Nuclear Power 

Facility, Appendix K (Atomics International, 1961, pp. 59, 65, 69, 72, 76). 

Containment Building 

The reactor was housed within a containment building that consisted of a steel shell with an inner 

lining of concrete. The steel shell was approximately 73 feet in inside diameter and 123 feet in overall 

 

3

 Terphenyls are aromatic hydrocarbons consisting of three benzene rings linked together with covalent bonds. Terphenyl 

is produced by a process that involves heating the benzene to about 600°C in the presence of a catalyst. 
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height. The shell was fabricated from A-201B steel and was 3/8 inch thick. Thicker plates were 

installed at the various penetrations points. The shell was lined with approximately 18 inches of 

concrete on the vertical sections above grade, and the dome was lined with concrete with a varying 

thickness from 18 inches at the lower edge to 6 inches at the top of the dome. The overall free volume 

within the building was approximately 300,000 cubic feet (Atomics International, 1965a, p. 3). 

The containment building was designed to withstand an internal pressure of 5 psig. At this pressure, 

the maximum leakage rate was 1% of the free volume per day. Atmospheric pressure was normally 

maintained within the building by means of supply and exhaust fans (Atomics International, 1965a, p. 

3). Airlocks were provided for access to the containment building. 

Fuel Handling System 

As described in the process and procedures defined in the Final Safeguards Summary Report for the 

Piqua Nuclear Power Facility, Appendix K, the specific fuel handling process was a remote process 

(Atomics International, 1961, pp. 302-305). The spent fuel rods, replaced during refueling operations, 

were transferred remotely to the fuel storage pool. The fuel storage pool was a separate storage 

system that contained an underwater rack for storing the spent fuel rods. Refueling was accomplished 

by working through the reactor top rotating shield, which contained a shield mounted on a circular 

bearing (to permit rotating around the vessel head flange) and a fuel removal port used during 

refueling operations. The fuel removal port could be located over any position in the reactor core. 

The fuel handling cask, which was a shielding device for transferring fuel into or out of the reactor, 

was equipped with a grappling and hoisting mechanism, a traveling bridge and carriage, and an 

emergency cooling system. Also involved in the fuel handling system was a fuel storage pool; the 

fuel storage pool was for the storage of spent fuel elements and consisted of a rack located at the 

bottom of the pool water. The storage holes in the rack were spaced to prevent fuel element 

criticalities (Atomics International, 1963). 

Heat Transfer System

The main heat transfer system consisted of a single coolant loop which contained two main coolant 

pumps, a superheater, a boiler, a surge tank, and the reactor vessel. Coolant flowed from the reactor 

vessel through the parallel connected main pumps, to the superheater and boiler, and back to the 

reactor vessel. The nominal flow rate through the system with a single pump in operation was about 

7,000 gpm, and with two pumps it was about 12,000 gpm. The flow rate was not less 220 gpm per 

megawatt of thermal power. The heat generated in the reactor core was removed by the circulation of 

the coolant in one downward pass through the core (Atomics International, 1965a, p. 5). 

Degasification System 

The degasification system was designed to remove waste gases and 8 pounds of water vapor per hour 

from the coolant. This system consisted of a degasifier tank which was designed to operate at a partial 

vacuum. The associated piping was provided to supply a sidestream of coolant from the main heat 

transfer system to the degasifier. There was a pressure reducing valve, located upstream of the 

degasifier, that controlled the main heat transfer system pressure and provided accessory control for 

maintaining vacuum in the degasifier. 
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Purification System 

The radiolytic decomposition of the coolant produced high-molecular-weight compounds (known as 

high boilers) which had to be removed from the coolant to maintain a fixed percentage of high boiler 

content. The coolant purification system continuously removed these compounds (which were 

subsequently transferred and treated/burned in the waste disposal system); thus, maintaining the 

desired high boiler content in the coolant in the main heat transfer system. The purification system 

also decontaminated the coolant, since most of the radioactivity in the coolant was removed with the 

high boilers (Atomics International, 1961, pp. 66-331). 

As described in Final Safeguards Summary Report for the Piqua Nuclear Power Facility, the 

purification system consisted of a distillation column, column feed heaters, condensers and still 

bottoms, and product receiver tanks. Purification system piping provided a sidestream of coolant 

from the coolant storage system and pressurization system into the column feed heaters. The 

purification system separated high boilers from the coolant and maintained the main heat transfer 

system coolant at a high boiler concentration of approximately 30% or less. The purification system 

was designed to process coolant at a flow rate of up to 1,000 pounds/hour. 

Aqueous Waste System 

The aqueous waste system consisted of a settling basin, waste holdup tanks, demineralizers, and 

interconnecting piping and sumps serving various process areas. The system was designed to gather 

and separate organic and particulate material from the aqueous wastes. Discharge from the settling 

basin could be stored in holdup tanks prior to disposal. Provisions were made for decontaminating the 

water, using an ion exchange technique, if radioactivity levels were above permissible levels. 

Waste Gas System 

The waste gas system consisted of a steam ejector, condensers, and holdup decay tanks. The system 

allowed gases to flow from the purification and degasification systems. The waste gases were 

processed through one of two decay tank banks; each bank contained eleven decay tanks, each having 

a diameter of 10 inches and a length of 15 feet. Four of the decay tanks in each bank were filled with 

activated carbon for processing waste gas prior to the gas being discharged through the stack. The 

decay tanks had a sufficient total capacity to delay the gases for 48 hours, giving ample time for 

radioactive decay of the process gases before exhausting into the atmosphere. The waste gas system 

provided storage and monitoring of all process gases prior to their release into the atmosphere through 

the stack. 

Organic Waste Disposal System 

The organic waste disposal system consisted of holdup tanks serving the purification system and a 

waste fired boiler designed to burn organic waste from the plant. Still bottoms from the purification 

system column were pumped into the decay tank and stored. The decay tank consisted of a 

compartmentalized vessel with seven compartments, each compartment having a capacity of 3,000 

gallons. The contents of each compartment were sampled prior to processing through the waste fired 

boiler. Organic wastes were burned subsequent to analysis. 
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Heating and Cooling System 

The heating and ventilation system provided for air circulation and heating or cooling in all buildings. 

It also provided for maintaining pressure differentials between various areas of the buildings, to 

ensure that air flow was always from non-contaminated areas into those areas having a higher 

potential for contamination. All ventilation air was filtered and monitored before being exhausted into 

the atmosphere. In the event of high radioactivity in the exhaust system, the reactor building had the 

ability to be automatically isolated, and the air would have been recirculated in a closed loop within 

the building. 

POMR Safety Features 

The POMR included many safety features that helped to ensure safe operation of the reactor and 

POMR personnel safety. Notable safety features included concrete for shielding, below-ground level 

components, process gases and waste monitoring, and automatic shutdown and air recirculation. 

Specific safety features include the following: 

• Concrete surrounded the reactor vessel (8’4” of concrete shielded or absorbed radiation emitted 

from the reactor core); 

• Continuous stack effluent monitors (including a particulate monitor and a gaseous activity 

monitor. Detection of airborne radioactivity levels above specified instrument radiation levels 

activated an alarm in the control room, and reactor building isolation devices would be activated 

upon detection of particulate activity);

• Personnel Monitors (Portal Monitor); 

• Exhaust Gas prefilter and absolute filter before going out of the 125 foot exhaust stack; 

• Below-ground-level components (including the core, associated piping, and organic auxiliary 

systems); 

• Solid concrete walls and partitions in the auxiliary building (to protect the environs from radiation 

from the piping and vessels in the building); 

• Sealed openings surrounding the reactor; 

• Containment shell (dome) was maintained at a negative pressure (with respect to the atmosphere) 

and retained any radioactive material released from the reactor vessel or process piping; 

• Airlocks used for containment shell 

• Noncorrosive properties of the organic coolant (corrosion of the fuel, piping, or other reactor 

equipment was unlikely); 

• Process gases were monitored (prior to entry into the train of decay tanks); 

• Potentially contaminated waste water was stored and monitored for safety determination (prior to 

being discharged from the plant); 

• Radiation monitors located at the exhaust stack (if excessive radioactivity was detected, the reactor 

building was automatically isolated and the air was recirculated within the building); 

• Automatic shutdown actions incorporated throughout the plant (in the event of off-normal 

conditions, the reactor would be shut down and/or isolated); 

• Fifteen remote area monitors (each detector operated a corresponding relay meter and a recorder in 

the control room to continuously indicate the radiation level from the area or equipment being 

monitored); 
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• Continuous monitoring of cooling water (via sample stream from the cooling water effluent taken 

prior to the effluent entering and mixing with industrial wastes); 

• Failed element location system (helped determine the presence of a failed fuel element in the core 

by monitoring the delayed neutron activity of the bulk outlet coolant);

• Annunciator point system (provided audio and visual indication of off-normal conditions);

• Fuel handling system (used cameras and periscopes to remotely view the operational areas); and

• Continuous air monitors (CAMs) (located throughout the plant).

Post-Operational Period 

In January 1966, when the reactor shut down due to a scram, several control rod drive problems were 

discovered; in particular, control rod 10 would not drop into place (Atomics International, 1966b). 

The reactor was opened and repairs were made to the control rods. In February, the investigation 

continued; some fuel elements were pulled (or were attempted to be pulled) in an effort to return them 

to their correct positions. However, many of the fuel elements failed to seat properly. On February 4, 

1966, one fuel element was removed and transferred to the fuel storage pool. This fuel element was 

found, using underwater television and photographs, to have buckles at three elevations and to have a 

“carbonaceous material” and a “fouling film” on its surface (Atomics International, 1966b). On 

February 24, 1966, after this fuel element was sent offsite for destructive analysis, the POMR facility 

suspended onsite recovery efforts and began to develop a plan for core unloading (Atomics 

International, 1966b). After discovering the buckled fuel element, operations were secured (through 

May 1, 1966) while the situation was evaluated by Atomics International and a plan for a path forward 

was developed. It should be noted that the buckling that was reported was observed in the outer 

process tube that formed the outer support surface of the fuel element and enclosed the fuel 

assemblies. A detailed report of the visual inspection of each fuel element included descriptions of 

each buckle and the surface appearance and location of the carbonaceous material (McCurnin, 1966). 

Nowhere in the documents reviewed was there a reference to a fuel cladding failure. 

5.2 Radiological Exposure Sources from POMR Operations 

The following subsections provide an overview of the internal and external exposure sources for the 

POMR class under evaluation. 

5.2.1 Internal Radiological Exposure Sources from POMR Operations 

ATTRIBUTION: Section 5.2.1 and its related subsections were completed by Daniel Mantooth, Dade 

Moeller and Associates, Inc. These conclusions were peer-reviewed by the individuals listed on the 

cover page. The rationales for all conclusions in this document are explained in the associated text.

The potential sources of alpha-emitting radionuclides at the POMR site included the uranium fuel 

itself, as well as transuranic radionuclides produced during the fission process while the reactor was 

running. However, in either case, these alpha-emitters would not be available for personnel exposures 

unless the aluminum cladding on the fuel was breached, allowing these radionuclides to contaminate 

the coolant and other parts of the system. The pre-demolition radiological survey performed in 1968, 

reported alpha contamination levels as “...non detectable during operation and less than 1 dpm/100 

cm2

 in final survey in all locations” (Wheelock, 1970, Appendix B). 
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The possibility of a cladding breach was discussed in one document (Ashley, 1964, p. 173) as being 

the cause of low levels of Xe-133 (a uranium fission product) detected in the process gas. However, 

the fact that the failed Fuel Element Location System (FELS) did not indicate such a failure led 

POMR site engineers to conclude that the xenon-133 was most likely a result of uranium 

contamination in the aluminum cladding. In any case, the amount of uranium involved would not 

have resulted in measureable exposures to POMR personnel. The absence of alpha emitting 

radionuclides was confirmed by radiological surveys conducted prior to the facility retirement 

(Wheelock, 1970, Appendix B, Table B-4). 

The primary sources of internal exposure at the POMR facility consisted of beta/gamma-emitting 

radionuclides from four sources: (1) activated impurities in the coolant, (2) activated corrosion 

products, (3) neutron recoil reactions with the aluminum cladding, and (4) tritium produced by ternary 

fission.4

5.2.1.1 Activated Impurities in the Coolant

The radionuclides that arise from activation of impurities in the coolant, as specified in Table 5-2, 

include sodium-24, phosphorus-32, sulfur-35, chlorine-38, manganese-56, arsenic-76, argon-41, and 

nitrogen-13 (Atomics International, 1961; Atomics International, 1965a; Atomics International, 

1964a), although argon-41 was believed to result from air in-leakage to the reactor vessel. The 

presence of nitrogen-13 was explained in the documentation as “...from the nitrogen cover gas” 

(Atomics International, 1965a). However, it is more likely that the carbon-14 was created from the 

carbon in the organic coolant from the carbon-12 (p, γ)

 13N reaction. The gaseous species argon-41 

and nitrogen-13 were removed by the coolant purification system, and after sufficient decay were 

exhausted via the main stack. Personnel had little opportunity for exposure to these gases during 

maintenance activities. Due to their short half-life, the majority of the activity in the system would 

have decayed prior to breaching coolant or other systems. POMR personnel outside the facility may 

have been exposed to low levels of these radionuclides during reactor plant operation. 

Ninety-six to ninety-eight percent of the non-gaseous radionuclides sodium-24, phosphorus-32, 

manganese-56, and arsenic-76 were thought to be collected along with the high-boiling organics from 

the coolant purification system. Seventy-eight percent of the sulfur-35 and chlorine-38 were collected 

in the same manner (Atomics International, 1961). Other systems where particulate activity could 

have been encountered include the in-vessel filters, coolant purification system, and the fuel storage 

pool purification filters. Operations that involved system maintenance or filter material replacement 

would have carried the potential for exposing personnel to these radionuclides. These radionuclides 

were present throughout the operational history of the reactor, but due to radioactive decay, the 

relative contributions would shift to longer-lived species during an outage period. Activated 

impurities were estimated to comprise 48% of the particulate radioactive species in the coolant 

(Atomics International, 1961). 

5.2.1.2 Activated Corrosion Products

 

4

 Attribution: The percentage reported for each source was calculated by Daniel Mantooth, Dade Moeller and Associates, 

Inc. from estimated concentration values found on page V-4 of Atomics International, 1961. 
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The primary radionuclide arising from the activation of corrosion products in the coolant is Mn-56. 

Since this radionuclide is a particulate, the discussion pertaining to locations in the reactor and 

mechanisms of exposure are identical to that presented for activated coolant impurities in Section 

5.2.1.1, above. Activated corrosion products (i.e., Mn-56) were estimated to comprise 13% of the 

total particulate activity (Atomics International, 1961). The radiological data for Mn-56 and potential 

sources are listed in Table 5-2. 

The reported method for radionuclide identification in the reactor coolant and effluents was by pulseheight analysis (gamma spectroscopy). This method would not have detected non-gamma emitting 

species such as tritium or carbon-14. The presence of these species was suggested in Final 

Safeguards Summary Report for the Piqua Nuclear Power Facility with calculated concentrations in 

coolant of 5.3E-04 μCi/cc and 1.4E-04 μCi/cc for tritium and carbon-14, respectively (Atomics 

International, 1961). Analytical results for tritium and carbon-14 in the coolant have not been found. 

Analyses of various process residues and water (Wheelock, 1970 Appendix B, Table B-2) in the fuel 

storage pool indicated tritium levels ranging from 7.5E-04 μCi/cc to 2.1E-02 μCi/cc. Results for 

carbon-14 ranged from 9E-05 μCi/cc to 2.1E-02 μCi/cc. Piqua Nuclear Power Facility Monthly 

Operating Report No. 24 reported that “bioassay performed for personnel working over the open 

reactor for tritium and net beta activity showed no positive result” (Atomics International, 1965b); 

these actual bioassay data have not been located. 

5.2.1.3 Neutron Recoil Reaction Products

Several particulate nuclides were thought to be produced by neutron recoil reactions with materials 

present in the aluminum fuel cladding (Atomics International, 1961; Atomics International, 1964c; 

Atomics International, 1965a; Atomics International, 1964a; Atomics International, 1965d). These 

include sodium-24 (also arises from impurities), magnesium-27, cobalt-58, and cobalt-60. [Note: Mg27 has a half-life of 9.5 minutes. Isotopes with half-lives less than 10 minutes provide negligible 

contribution to internal dose. Thus, magnesium-27 will not be included as a radionuclide of 

concern.]5 Since these radionuclides are particulates, the discussion pertaining to locations in the 

reactor and mechanisms of exposure is identical to that previously presented for activated coolant 

impurities in Section 5.2.1.1. Radionuclides arising from recoil reactions were estimated to comprise 

39% of the total particulate activity (Atomics International, 1961). 

5.2.1.4 Fission Products

Potential fission products such as xenon-133m, xenon -135, krypton-85m, krypton-85, and krypton-87 

were detected in the process gas, but only at levels on the order of a few disintegrations per minute per 

gram (dpm/g). Considering the low activity levels and the fact that these isotopes are noble gases and 

are only significant from an external exposure standpoint, they will not be considered further as a 

source of internal dose. 

The Final Safeguards Summary Report for the Piqua Nuclear Power Facility proposed that the 

presence of tritium in the coolant was primarily a result of the ternary fission in the fuel (Atomics 

International, 1961, p. V-6). Twenty-five percent of the tritium created was assumed to enter the 

 

5

 Attribution: This decision was based on comments from the ORAU Team’s DOE Site Principal Internal Dosimetrist, 

Elizabeth Brackett, MJW Corporation. 
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coolant by diffusion or recoil processes, which would have resulted in an estimated equilibrium 

concentration of 0.21 μCi/cm3 at full power. This concentration is 10 times greater than the maximum 

value measured in process residues reported above in Section 5.2.1.2. which were analyses of various 

process residues in systems and water in the fuel storage pool. Analytical results for tritium in the 

coolant nor environmental monitoring data been located. However, an analysis for tritium in the 

carbonaceous material discovered in the coolant resulted in a level of 3.9 μCi/g (Atomics 

International, 1966b). Piqua Nuclear Power Facility Monthly Operating Report No. 24 mentioned 

that “bioassay performed for personnel working over the open reactor for tritium and net beta activity 

showed no positive result” (Atomics International, 1965b). However, no bioassay data have been 

located. 

Table 5-2: Radionuclides of Concern for Internal Exposure

Radionuclide Half-life* Primary Radiations 

and Energies (Mev)* Potential Source Reactions

Na-24 15 hr β- 4.17 

γ 1.369 

Coolant Impurities: 23Na(n,γ)

24Na* 

Recoil: 27Al(n,α)

24Na**

P-32 14.28 d β- 1.71 Coolant Impurities: 32S(n,p)32P* 

S-35 87.9 d β- 0.167 Coolant Impurities: 37Cl(d,a)35S* 

Coolant Impurities: 34S(n,γ)

35S* 

Cl-38 37.29 min β- 4.91 

γ 1.60 

Coolant Impurities: 37Cl(n,γ)

38Cl* 

Mn-56 2.58 hr β- 2.85 

γ 0.847 

Corrosion/Impurities: 55Mn(n,γ)

56Mn* 

Co-58 70.88 d β+ 0.474 

γ 0.810 (99%) 

Recoil: 58Ni(n, p)58Co** 

Co-60 5.26 yr β- 0.314 (99%) 

γ 1.173 (100%) 

 1.133 (100%) 

Recoil: 60Ni (n,p)60Co* 

As-76 26.4 hr β- 2.97 

γ 0.559 (43%) 

 0.957(6%) 

 1.22 (5%) 

 ≤2.1 (2%) 

Coolant Impurities: 75As(n,γ)

76As* 

Tritium 12.3 yr β- 0.0186 Coolant Activation 2H(n,γ)

3

H* 

C-14 5730 yr β- 0.156 Coolant Activation 13C(n,γ)

14C* 

Notes: 

* Data are from Radiological Health Handbook (Radiological Health Handbook, 1970).

** Data are from Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 1961). 

5.2.2 External Radiological Exposure Sources from POMR Operations 

ATTRIBUTION: Section 5.2.2 and its related subsections were completed by Louise Buker, Oak Ridge 

Associated Universities (ORAU) and Roger Halsey, Oak Ridge Associated Universities (ORAU). 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

Workers at the POMR facility were potentially exposed to external photon, beta, and/or neutron 

radiation from activities associated with maintaining and operating the reactor. Potential exposure 
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sources included radioactive materials from the operation and maintenance of the nuclear reactor and 

radioactive materials in the form of calibration sources. 

5.2.2.1 Photon

Some POMR radiological operations potentially involved gamma and X-ray photon radiation fields. 

The potential photon exposure sources would have included the following: 

• Gamma-emitting fission and/or activation products resulting from the reactor operations, 

• Bremsstrahlung radiation from various beta-emitting radionuclides, and 

• Calibration sources of cobalt, cesium, and other miscellaneous radionuclides (Geiger, 1969). 

During operations, the radiation levels from the coolant lines and filters would have resulted from 

activation of corrosion products and other impurities in the coolant, as well as from recoil products 

from the core. The short lived gamma-emitting isotopes that have a half-life of 10 minutes or less 

(magnesium-27 and nitrogen-13) would be responsible for the majority of the radiation from the 

coolant lines (Atomics International, 1965a, p. 265). During reactor operations, the highest radiation 

levels in the plant were at the degasifier filters, F 2A and F 2B (Atomics International, 1965a, pp. 265 

-270). 

5.2.2.2 Beta

There was a potential for beta particle-emitting source term during shutdowns, maintenance, refueling, 

and when fuel was removed. Beta radiation could have resulted from activation and fission products. 

During normal operations, fission and activation products would have been located within the core 

and within the various shielding surrounding the core. 

NIOSH located documentation explaining the beta radiation levels during maintenance activities. The 

documentation stated that “beta radiation is the primary type of activity encountered when systems are 

opened.” It further stated that through the entire test program, as a result of the low dose rates present, 

it was unnecessary to establish work time limits for personnel performing maintenance work on 

exposed system components (Atomics International, 1965a, p. 275). 

The potential for beta exposure from fission and activation products was known and documented prior 

to plant startup (Atomics International, 1963, p. 83). 

5.2.2.3 Neutron

There was a low potential for neutron radiation exposure associated with POMR operations. Neutron 

exposures could have occurred, as a result of the fission process, from operating the reactor from 1963 

to 1966. In addition, there was some potential for neutron exposures from a plutonium-beryllium 

(PuBe) neutron source for those personnel who calibrated the neutron survey instruments (Personal 

Communication, 2009g). The potential personnel exposures from this type of check source is 

considered to be encompassed in the assessment of reactor-related neutron exposures included in this 

evaluation. The rate of neutron generation from reactor operations was based on the rate of fission 

which would have been directly proportional to the reactor power level. An interviewee mentioned 
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that there was a plutonium-beryllium neutron calibration source used for checking the long counter6

, 

which was an energy-independent neutron measuring instrument (Personal Communication, 2009g). 

The source of the neutron emissions and potential worker exposures was minimized as a result of the 

design of the POMR, which controlled the operationally-related neutron exposure sources. The 

POMR was well-shielded by concrete that surrounded the reactor vessel, below-ground-level 

components, and sealed openings that surrounded the reactor (Unknown author, unknown date-b). 

6.0 Summary of Available Monitoring Data for the Class Evaluated 

by NIOSH 

The following subsections provide an overview of the state of the available internal and external 

monitoring data for the POMR class under evaluation. 

6.1 Available POMR Internal Monitoring Data 

ATTRIBUTION: Section 6.1 was completed by Daniel Mantooth, Dade Moeller and Associates, Inc. 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

No primary source bioassay monitoring data (e.g., original dosimetry records) were located for POMR 

personnel for the NIOSH-evaluated period. One report, Piqua Nuclear Power Facility Monthly 

Operating Report No. 24 mentions that urine samples were collected for personnel “...working over 

the open reactor,” but the bioassay showed no positive result tritium and net beta activity. One former 

employee reported that bioassays were performed annually (Personal Communication, 2009h). 

 

Maximum and occasionally average stack effluent activity concentrations were reported in monthly, 

quarterly, and semiannual reports (Atomics International, 1964b; Atomics International, unknown 

date; Atomics International, 1965b; Atomics International, 1965d; Atomics International, 1964c). The 

concentrations reported in the reports are consistent with the calculated estimates provided in the 

Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 

1961, Table V-3). 

Maximum and average air activity concentrations for environmental and onsite samples are provided 

in three reports—Piqua Nuclear Power Facility Monthly Operating Report No. 24; Piqua Nuclear 

Power Facility Monthly Operating Report No. 26; and Piqua Nuclear Power Facility Monthly 

Operating Report No. 19. The location of these samples could not be determined from the available 

information. The concentrations reported are substantially less than the maximum permissible 

 

6

 A long counter is a neutron detector that is designed to measure all neutrons accurately. It has a flat response over a 

broad range of neutron energies. At its core, it has a BF3 proportional counter, able to discriminate neutrons against 

gamma activity by use of the neutron-alpha reaction on boron-10. This core is surrounded by a cylinder of paraffin. Fast 

neutrons are moderated by the paraffin and are then captured by the boron-10 in the BF3 gas. Discriminators are used to 

accept only the signals from the alpha and lithium-7 recoil products. The detectors are energy independent, typically in the 

range from 10 keV to 5 MeV and are shielded to provide a highly directional response (Cember, 1969). 
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concentration values shown in Table V-3 of Final Safeguards Summary Report for the Piqua Nuclear 

Power Facility (Atomics International, 1961).

No primary source workplace air or breathing zone air sample data (e.g., original records) were 

located. However, there were instances in which levels were discussed in relative terms, (e.g., “...no 

personal contaminations or inhalations” (Atomics International, 1965b); “Airborne activity in the 

containment building has not exceeded that normally observed from natural background” (Atomics 

International, unknown date)) 

6.2 Available POMR External Monitoring Data 

ATTRIBUTION: Section 6.2 was completed by Louise Buker, Oak Ridge Associated Universities 

(ORAU) and Roger Halsey, Oak Ridge Associated Universities (ORAU). These conclusions were 

peer-reviewed by the individuals listed on the cover page. The rationales for all conclusions in this 

document are explained in the associated text. 

NIOSH has not located any primary source external dosimetry records (e.g., original dosimetry 

records) for the Piqua workers. External data for the POMR facility that have been found are in the 

form of exposure summary reports and monthly and semiannual operational progress reports for the 

years 1963 through 1966. The historical Piqua documentation demonstrates knowledge of potential 

workplace external radiation hazards, applicable radiation exposure guidelines, methods for limiting 

worker exposure, and radiation monitoring and dosimetry capabilities. 

Photon

The most complete, available data consist of summary dosimetry data reported annually to the AEC 

offices for Chicago Operations and Idaho Operations. These summary data provide evidence that 

external monitoring was performed for the Piqua site. Table 6-1 summarizes the dosimeter results that 

were sent annually to the AEC Chicago Operations Office. 

Table 6-1: Piqua Summary Annual Whole-Body Penetrating Exposures 

Year No. Identified as 

NOT Monitored 

No. Identified as 

Monitored 

No. Identified with 

0-1 rem 

No. Identified with 

1-2 rem 

1963 0 42 42 - 

1964 0 46 46 - 

1965 0 47 47 - 

1966 0 50 42 8 

Notes: 

- indicates that no exposures were reported in this range. 

Information for 1963 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1963 (AEC, 1963, p. 34). 

Information for 1964 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1964 (AEC, 1964, p. 18). 

Information for 1965 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1965 (AEC, 1965a, p. 27). 

Information for 1966 is from Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1966 (AEC, 1966, p. 23). 
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The summary data that were reported to the AEC Idaho Operations Office for the year 1963 indicated 

that 31 Atomics International employees worked at the Piqua facility with the City of Piqua workers, 

performing the same types of jobs. The Atomic International employees’ dosimeters indicated that 

they received 0 gamma, 0 beta, and 0 neutron dose for the year 1963 (City of Piqua, 1964). One 

worker did have detectable gamma of 60 mrem for the year. Individual data for the City of Piqua 

workers are unavailable at this time. 

Monthly and semiannual operational reports indicate that R. S. Landauer Jr. and Company was the 

dosimetry provider. However, individual dosimetry records are not currently available; thus, the 

information that NIOSH has available to supplement the data provided in Table 6-1 consists of 

operational progress reports from 1963 through 1966. These reports include references to ambient 

radiation levels and routine survey results. Table 6-2 provides a summary of the available data from 

semiannual and monthly reports that were sent to the Chicago AEC Operations Office. 

Table 6-2: Piqua Semiannual and Monthly Report Results 

Table 6-2 and its corresponding notes span two pages. 

Timeframe Results 

Report Type 

and Number Begin End 

Number of 

Positive 

Results 

Maximum 

Result 

Notes from the Radiation Monitoring 

Section of the reports 

Semiannual-2 

(Atomics 

International, 

1963) 

01/01/63 06/30/63 Unknown Unknown No radiation measurement mentioned 

(Atomics International, 1963) 

Semiannual-3 

(Atomics 

International, 

1964d) 

07/01/63 12/31/63 Unknown Unknown No radiation measurement mentioned 

(Atomics International, 1964d) 

Semiannual-4 

(Atomics 

International, 

1964a) 

01/01/64 06/30/64 Unknown Unknown Radiation levels measured at eighteen 

locations while operating at full power (45.5 

MWt.) 

No neutrons detected from the biological 

shield or above the reactor at the 100’ level 

(Atomics International, 1964a, p. 82) 

Location OC-104 <0.5 mrem/hr neutrons 

Location OC-101, 14" line in the P-1A room 

<0.5 mrem/hr neutrons (Atomics 

International, 1964a, p. 83) 

During shutdown in June, the following 

measurements were taken: 

• 1 R/hr measured at in-core filter that had 

been removed (Atomics International, 

1964a, p. 17) 
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Table 6-2: Piqua Semiannual and Monthly Report Results 

Table 6-2 and its corresponding notes span two pages. 

Timeframe Results 

Report Type 

and Number Begin End 

Number of 

Positive 

Results 

Maximum 

Result 

Notes from the Radiation Monitoring 

Section of the reports 

• Filter media measured < 20 mrad/hr beta 

- gamma, including 5 mR/h gamma 

(Atomics International, 1964a, p. 17) 

Semiannual-5 

(Atomics 

International, 

1965e) 

07/01/64 12/31/64 Unknown Unknown 2-4 mR/h surface of main coolant piping 

(Atomics International, 1965e, pp. 61-63) 

Monthly-19 

Atomics 

International, 

1964c) 

10/01/64 10/31/64 10 30 mrem N/A 

Semiannual-6 

(Atomics 

International, 

1965c) 

01/01/65 06/30/65 Unknown 120 mrem During the month of April, this max 

estimated dose was received during 

inspection from non-routine removal of fuel 

elements (Atomics International, 1965c, pp. 

76-77) 

700mR/h degasification filter, attributed to 

Mg-27 and Na-24 from the 6 new fuel 

elements (Atomics International, 1965c, pp. 

76-77) 

Monthly-24 

Atomics 

International, 

1965b) 

03/01/65 03/31/65 4 80 mrem N/A 

Simiannual-7 

(Atomics 

International, 

1966a) 

07/01/65 01/13/66 Unknown Unknown No section for plant radiation levels; some 

discussion of dose rates for a degasifier 

(Atomics International, 1966a, p. 66) 

Monthly-26 

(Atomics 

International, 

1965d) 

05/01/65 05/31/65 2 290 mrem Dose received by HP personnel during 

reactor shutdown monitoring and calibration 

of HP instruments (Atomics International, 

1965d, p. 20) 

Semiannual-8 

(Atomics 

International, 

1966b) 

01/01/66 06/30/66 Unknown Unknown 5 mR/h beta and gamma; sample of coke 

surface (Atomics International, 1966b, p. 

103) 

Notes: 

 “Unknown” indicates that information was not provided or available. 

For 1963, the available operational reports—Piqua Nuclear Power Facility Operations Analysis 

Program Progress Report No. 2 (Fiscal Year 1963) and Piqua Nuclear Power Facility Operations 

Analysis Program Progress Report No. 3 (July 1, 1963-December 31, 1963)—did not discuss the 

highest dose for POMR personnel (Atomics International, 1963; Atomics International, 1964d). 

However, there is a 1963 report from the City of Piqua for 31 Atomics International workers that 
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worked at the site; this report lists 0 gamma, 0 beta, and 0 neutron dose for thirty of the employees. 

For one employee, the report lists 60 mrem gamma (City of Piqua, 1964). In addition, there is a 

summary of quarterly cycle data that was submitted to the AEC for one Piqua employee for the years 

1963 through 1965, where it was reported that he received 0 photon, 0 beta, and 0 neutron dose for 

each of those years (AEC, 1965b). 

For 1964, Piqua Nuclear Power Facility Monthly Operating Report No. 19 reported only one 

maximum exposure of 30 mrem for the month of October, and 10 readings that were greater than 

detectable levels for deep dose (Atomics International, 1964c). The semiannual operational reports 

did not mention the highest exposure for the period. The reports include summaries of survey 

monitoring data that provide dose rates at certain areas of the plant for beta, gamma, and neutrons. 

Additionally, Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 4 

(January 1, 1964-June 30, 1964) provides a summary of the radiation levels associated with the 

megawatt thermal energy of the reactor from January through May of 1964 for survey monitoring 

dose rates at the “waste gas activity” location, the “out of core bypass filter activity” location, and the 

“main coolant superheater outlet line” location. The highest level was approximately 400 mR/hr at 

the “out of core bypass filter” location; the “waste gas activity” and the “main coolant superheater 

outlet line” measurements were less than 30 mR/hr (Atomics International, 1964a, p. 80). 

For 1965, the available operational reports— Piqua Nuclear Power Facility Reactor Operations 

Analysis Program Semiannual Progress Report No. 6 (January 1-June 30, 1965) and Piqua Nuclear 

Power Facility Monthly Operating Report No. 24—report that there were maximum doses reported of 

80 mrem in March; 120 mrem in April during a non-routine inspection of fuel elements; and 290 

mrem in May, received during calibration activities during a reactor shutdown period (Atomics 

International, 1965c; Atomics International, 1965b). 

For 1966, Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress 

Report No. 8 (January 1-June 30, 1966) did not report the maximum external exposure at the site 

(Atomics International, 1966b). However, the Site Summary reports identified in Table 6-1 did report 

findings that exceeded the 1 rem value (but less than the 2 rem value) for eight workers. 

Beta

NIOSH has not located a complete set of data covering beta or shallow measurements. The monthly 

and semiannual reports have reported beta or shallow measurements occasionally with gamma 

measurements. The annual summary reports to the AEC Chicago Operations (the whole-body 

penetrating exposures are provided in Table 6-1) do not provide a beta dose because only whole-body 

penetrating exposure ranges were being reported. 

However, the external dosimetry for plant workers (film badges provided by R. S. Landauer Jr. and 

Company) included a shallow dose component (Atomics International, 1965b, p. 23). In the summary 

data that were reported to the AEC Idaho Operations Office for 1963, the Atomic International 

employees’ dosimeters indicated that 30 of the 31 individuals monitored received 0 beta dose for the 

year. One individual had a reported dose of 20 mrem beta (City of Piqua, 1964). 
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Survey data reported in the annual and semiannual reports typically were reported as β-γ (betagamma). There were couple of instances where the gamma exposure rate was distinguished from the 

beta-gamma total. These instances described scenarios with the potential for personnel exposure. The 

first instance described a measurement of an in-core filter that had been removed. It had a reading of 

20 mrad/hr beta-gamma including 5 mr/hr gamma (Atomics International, 1964a, p. 17). The second 

instance described a survey that was taken inside the purification column and overheads cooler that 

had been opened for maintenance. The highest reading found at this location was 10 mrads/hr betagamma, including 0.5 mrad/hr gamma at 1” from the contaminated surface (Atomics International, 

1964c, p. 19). 

Neutron

There is no complete set of data that explicitly covers neutron dose. However, based on the available 

summary data, the individual data reports (reported to the AEC as whole-body radiation exposures to 

penetrating radiation) (AEC, 1963; AEC, 1964; AEC, 1965a; AEC, 1966) would likely have included 

neutron dose, had any been detected. In the summary data that were reported to the AEC Idaho 

Operations Office for 1963, the 31 Atomic International employees’ dosimeters indicated that they 

received 0 neutron dose that year (City of Piqua, 1964). 

During an interview, a former POMR health physics technician stated that neutron surveys were 

routine and performed initially each time the power level was raised (Personal Communication, 

2009g). One set of neutron results collected in 1964 was reported in a semiannual report (Atomics 

International, 1964a p. 83). Three locations were listed as having no neutrons, while two others were 

listed as having <0.5 mrem/hr neutrons and had a corresponding gamma dose rate. The survey was 

taken while the plant was operating at full power (45.5 MWt). 

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by 

NIOSH 

The feasibility determination for the class of employees under evaluation in this report is governed by 

both EEOICPA and 42 C.F.R. § 83.13(c)(1). Under that Act and rule, NIOSH must establish whether 

or not it has access to sufficient information either to estimate the maximum radiation dose for every 

type of cancer for which radiation doses are reconstructed that could have been incurred under 

plausible circumstances by any member of the class, or to estimate the radiation doses to members of 

the class more precisely than a maximum dose estimate. If NIOSH has access to sufficient 

information for either case, NIOSH would then determine that it would be feasible to conduct dose 

reconstructions. 

In determining feasibility, NIOSH begins by evaluating whether current or completed NIOSH dose 

reconstructions demonstrate the feasibility of estimating with sufficient accuracy the potential 

radiation exposures of the class. If the conclusion is one of infeasibility, NIOSH systematically 

evaluates the sufficiency of different types of monitoring data, process and source or source term data, 

which together or individually might assure that NIOSH can estimate either the maximum doses that 

members of the class might have incurred, or more precise quantities that reflect the variability of 
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exposures experienced by groups or individual members of the class. This approach is discussed in 

OCAS’s SEC Petition Evaluation Internal Procedures, which are available at 

http://www.cdc.gov/niosh/ocas. The next four major subsections of this Evaluation Report examine: 

• The sufficiency and reliability of the available data. (Section 7.1) 

• The feasibility of reconstructing internal radiation doses. (Section 7.2) 

• The feasibility of reconstructing external radiation doses. (Section 7.3) 

• The bases for petition SEC-00126 as submitted by the petitioner. (Section 7.4) 

7.1 Pedigree of POMR Data 

This subsection answers questions that need to be asked before performing a feasibility evaluation. 

Data Pedigree addresses the background, history, and origin of the data. It requires looking at site 

methodologies that may have changed over time; primary versus secondary data sources and whether 

they match; and whether data are internally consistent. All these issues form the bedrock of the 

researcher’s confidence and later conclusions about the data’s quality, credibility, reliability, 

representativeness, and sufficiency for determining the feasibility of dose reconstruction. 

Available data for the POMR facility are in the form of summary reports, including monthly, 

semiannual, and annual reports for the years 1963 through 1966. The types of monitoring data 

reported in the monthly and semiannual reports are inconsistent from report to report. If we were to 

only use these reports, there would be some information gaps for all years of operation, and NIOSH 

does not have access to the individual, hardcopy monitoring data. 

7.1.1 Internal Monitoring Data Pedigree Review 

ATTRIBUTION: Section 7.1.1 was completed by Daniel Mantooth, Dade Moeller and Associates, Inc. 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

No original records for bioassay or area air samples, or (surface contamination) data records have 

been located. There are several cases where air sampling results (reported as maximum, minimum, 

and average) for stack effluent and environmental activity levels are reported in summary documents. 

Other documents report these parameters and workplace air levels subjectively (i.e., < MPC). A data 

quality review cannot be performed on the summary data that was reported to the AEC without the 

actual data records. The summary documents reviewed consist of various monthly, quarterly, and 

semiannual reports of reactor plant conditions that collectively cover the entire thirty-two month 

period in which the reactor operated. Although this information is not the original data, the 

information contained within these documents, when taken together, indicates that POMR radiological 

conditions were benign with respect to internal dose; an indication supported by the reports of former 

POMR personnel interviewed during the course of this evaluation report (Personal Communication, 

2009a; Personal Communication, 2009e; Personal Communication, 2009g). Collectively, every 
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source reviewed and interview conducted leads to the conclusion that surface and airborne 

contamination levels and radionuclides present during POMR operations were unlikely to have 

resulted in measurable internal dose. 

7.1.2 External Monitoring Data Pedigree Review 

ATTRIBUTION: Section 7.1.2 was completed by Louise Buker, Oak Ridge Associated Universities 

(ORAU). These conclusions were peer-reviewed by the individuals listed on the cover page. The 

rationales for all conclusions in this document are explained in the associated text. 

Although external monitoring did occur at POMR, NIOSH has not yet been able to locate the original 

data records. The only external monitoring data available to NIOSH consist of summary dosimetry 

data reported annually to the AEC offices of Chicago Operations and Idaho Operations. Without the 

actual data records, a data quality review cannot be performed on the summary data that were reported 

to the AEC. 

However, there is corroborating evidence that supports NIOSH’s ability to bound the external dose for 

the class evaluated in this report. A 1963 summary report to the AEC Idaho Operations Office 

indicates that 30 Atomics International workers who worked at the POMR facility received 0 gamma, 

0 beta, and 0 neutron dose for the year 1963, and one Atomics International worker received 60 mrem 

gamma, 20 mrem beta and 0 mrem neutron (City of Piqua, 1964). In addition, former POMR 

personnel interviewed during the course of this evaluation report provided information about the 

POMR facility’s radiation protection practices that coincides with the available summary data. 

Information obtained from interviews with former POMR facility personnel indicates that (1) 

personnel working at the site were issued monthly or quarterly film badges, and (2) area surveys were 

conducted on a routine and/or as-needed basis. Also, the monitoring was conducted for beta, gamma, 

and neutron radiation. NIOSH has compared the actual result information (as relayed in Section 6.2, 

and contained in Table 6-2, of this report) to the maximum values in the ranges of doses listed in the 

summary data reports for the respective operational years. This comparison supports NIOSH’s 

conclusion that the summary report data provided in Table 6-1 can be used as a bounding estimate for 

the external doses for the evaluated class. 

7.2 Evaluation of Bounding Internal Radiation Doses at POMR 

ATTRIBUTION: Section 7.2 and its related subsections were completed by Daniel Mantooth, Dade 

Moeller and Associates, Inc. These conclusions were peer-reviewed by the individuals listed on the 

cover page. The rationales for all conclusions in this document are explained in the associated text. 

The principal source of internal radiation doses for members of the class under evaluation was 

exposure to beta/gamma emitting radionuclides from four sources: (1) activated impurities in the 

coolant, (2) activated corrosion products, (3) neutron recoil reactions with the aluminum cladding, and 

(4) tritium produced by ternary fission. The following subsections address the ability to bound 

internal doses, methods for bounding doses, and the feasibility of internal dose reconstruction. 
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7.2.1 Evaluation of Bounding Process-Related Internal Doses 

NIOSH was unable to locate any urinalysis data, lung-count data, or any other type of bioassay data. 

Detailed below is a summary of the particulate airborne activity information available for 

reconstructing the process-related internal doses of members of the class under evaluation. 

As discussed previously, particulate radionuclides consisting of sodium-24, phosphorus-32, sulfur-35, 

chlorine-38, manganese-56, arsenic-76, magnesium-27, cobalt-58, and cobalt-60 were detected in the 

organic coolant, in in-vessel filters, and in filters associated with the coolant purifications system, the 

degasifications system, the fuel storage pool-purification system, and the waste fired boiler. It was 

proposed that tritium and carbon-14 would be present in the coolant at levels up to 0.21 μCi/cc and 

1.4e-04 μCi/cc, respectively (Atomics International, 1961). Airborne activity, if present, would have 

been generated by maintenance activities on these systems. As mentioned previously, no work area 

air sampling data have been located that could be used to support establishing bounding estimates of 

the airborne activity. 

The Final Safeguards Summary Report for the Piqua Nuclear Power Facility (Atomics International, 

1961) provides calculations demonstrating that airborne levels would be less than the MPC specified 

in the 1960 regulations (current at the time) (10 C.F.R. pt. 20). Summary documents reported air 

sampling results in relative terms (i.e., “...no personal contaminations or inhalations” and “...airborne 

levels in containment building were less than background”) (Atomics International, 1964a; Atomics 

International, 1965a; Atomics International, unknown date; Atomics International, 1965b) that 

corroborate the safety analysis estimation. Because the reactor was a closed-loop primary system, and 

the coolant became a solid at room temperature, it was unlikely that personal contamination and 

inhalation was a factor.7

 Air monitoring capability is described in Compilation of Piqua Nuclear 

Power Facility Operating Limits and Controls and Post-Critical Operational Tests, and confirms that 

the air sampling equipment (with the exception of tritium and carbon-14) could detect total betagamma activity at the MPC level for cobalt-60 (9E-9μCi/cm3

) for restricted areas and would alarm 

before activity approached hazardous levels (Atomics International, 1965a). The reported sensitivity 

for workplace air monitors for particulate activity was 1E-12 μCi/cm3

 (Atomics International, 1965a), 

or nearly 3 orders of magnitude less than the MPC. The presence of workplace air samplers, their 

capabilities, and consistently low airborne activity levels were also confirmed by former Piqua 

personnel (Personal Communication, 2009a; Personal Communication, 2009e; Personal 

Communication, 2009g). 

There were no incidents reported that led to significant personnel contamination. As an example, 

Piqua Nuclear Power Facility Reactor Operations Analysis Program Semiannual Progress Report 

No. 6 (January 1-June 30, 1965) mentioned that the soot collection bag from the waste fired boiler 

came loose and released contaminated soot into the auxiliary building. Total surface contamination 

levels of 1,000 dpm/44 in2

 (~400 dpm/100 cm2) were measured (Atomics International 1965c, p. 77). 

 

7

 Attribution: This conclusion was made by Daniel Mantooth, Dade Moeller and Associates, Inc. and is based on five 

interviews that referenced a solid coolant (Personal Communication, 2009a; Personal Communication, 2009c; Personal 

Communication, 2009d; Personal Communication, 2009f; Personal Communication, 2009g) 
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No personnel contaminations or inhalations occurred during this event. The radioisotopes involved 

were not specified, but it can be assumed that they were comprised of those described above and listed 

in Table 5-2. Based on its research to date, NIOSH has not located air sampling data sufficient by 

themselves to support bounding process-related internal dose. However, the information discussed 

above and further evaluated below, supports the conclusion that airborne particulate activity in process 

areas was not likely to exceed the MPCs. The bounding process-related internal doses from the 

particulate radionuclides can be obtained using the MPC values. Intakes of tritium and carbon-14 are 

calculated using the maximum ratio of the reported or measured activities in the coolant or process 

residues, respectively (see Section 5.2.1.2) to the total reported particulate activity. The resulting ratio 

is then applied to the expected process airborne level (i.e., MPC) to derive the bounding value for 

tritium and carbon-14. 

7.2.2 Evaluation of Bounding Ambient Environmental Internal Doses 

Ambient environmental internal doses could have resulted from the inhalation of radionuclides 

exhausted from the reactor plant. Ambient air particulate levels were monitored in stack effluents at 

both on- and off-site (so-called “environmental samples”) locations. Data for stack particulates have 

not been located, but several reports indicate that stack effluents were consistently “...less than the 

MPC” (Atomics International, unknown date; Atomics International, 1964a; Atomics International, 

1964b; Atomics International, 1964c; Atomics International, 1965c). In addition, the stack 

monitoring system was reported to have sufficient sensitivity to detect particulate activity equal to 3E10 μCi/cm3

 of iodine-131 (Atomics International, 1965a), which is equivalent to the unrestricted MPC 

for cobalt-60, the limiting radionuclide of concern at the POMR facility. Since the maximum beta 

energies for iodine-131 and cobalt-60 are on the same order (i.e., 0.606 and 0.318), it can be assumed 

that the monitoring instrument sensitivity would be similar. Alarms would sound in the control room 

when activity exceeded the MPC level. The information regarding stack effluent values and 

monitoring coincides with the estimated values provided in the Final Safeguards Summary Report for 

the Piqua Nuclear Power Facility (Atomics International, 1961). In the case of onsite and offsite 

monitoring, several monthly reports provide results of this sampling (Atomics International, 1964c; 

Atomics International, 1965b; Atomics International, 1965d), with all indicating a total particulate 

activity well below the outside operational (unrestricted area) MPC values.

Based on its research to date, NIOSH has located limited air sampling data to support bounding 

ambient environmental internal dose. However, the information discussed previously supports 

NIOSH’s conclusion that airborne activity levels outside the process areas and the plant grounds can 

be bounded by application of operational internal dose assessment methods. 

7.2.3 Methods for Bounding Internal Dose at POMR 

7.2.3.1 Methods for Bounding Operational Period Internal Dose

The method for bounding the operational period internal dose is based on the conclusion that the 

airborne activity in operational areas would not have exceeded the lowest applicable MPC without 

being detected by the existing monitoring system and noted in the routine summary reports. This 

conclusion is based on information in the SRDB and interviews with former POMR facility personnel 

indicating that the POMR facility had, and used, workplace alarming airborne monitoring equipment 
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capable of detecting airborne particulate activity to the MPC levels in effect at the time (as noted in 10 

C.F.R. pt. 20, Appendix B Table II). Typically, the alarm on air monitors is set at some percentage of 

the applicable limit for the most restrictive radionuclide (i.e., lowest MPC). For the purposes of 

bounding the operational period internal dose, it will be assumed that the alarms were set to trip at the 

MPC for the most restrictive radionuclide. The radionuclide with the lowest MPC listed in Table 5-2 

is chlorine-38 (3E-09 μCi/cm3

). However, due to its short half-life (37.3 min), it is likely that 

chlorine-38 levels could have largely decayed away before sufficient activity would be collected by 

the air monitor to cause an alarm. The radionuclide with the next lowest MPC is cobalt-60 (9E-09 

μCi/cm3

); with a half-life of over 5 years, sufficient activity would collect to reach alarm levels if 

elevated airborne activity was present. If the total beta/gamma airborne activity did not exceed the 

MPC for cobalt-60 (9E-09 μCi/cm3

), and the total particulate activity consisted of 48% impurities, 

13% corrosion products, and 39% recoil products (as noted in Section 5.2.1), the total air activity 

would be comprised of 4E-09 μCi/cc impurities, 1E-09 μCi/cc corrosion products, and 4E-09 μCi/cc 

recoil products. These concentrations can be used to assess the radionuclide exposures in each 

applicable category (Table 7-1). The radionuclide that represents the maximum internal dose 

contributor in each category can be evaluated on a case-by-case basis, and the categories combined to 

establish a composite dose estimate. The application of this method supports NIOSH’s ability to 

bound the internal dose from exposure to these radionuclides for the evaluated class. 

To calculate the airborne levels of tritium and carbon-14 in process areas, ratios were calculated 

between the maximum expected or reported activity of these two radionuclides to the total activity in 

the coolant and residues from other radionuclides (activation, corrosion, and recoil), as described 

previously. Information obtained from several documents (Atomics International, 1964b; Atomics 

International, 1964a; Atomics International, 1965d; Atomics International, 1965b) provide data 

indicating an average coolant activity over a wide range of conditions of 1.1e-02 μCi/cm3

. The 

greatest tritium activity discovered in the available on the SRDB documentation is the estimate 

provided by Final Safeguards Summary Report for the Piqua Nuclear Power Facility, or 0.21 

μCi/cm3. The ratio of this tritium activity to the total activity would be 19.1 (0.21 μCi/cm3 ÷ 1.1e-02 

μCi/cm3

 = 19.1). Since the maximum airborne activity is not expected to exceed the MPC for cobalt60 (9E-09 μCi/cm3

), the tritium airborne concentration should not exceed 1.7E-07 μCi/cm3 (19.1 x 

9E-09 μCi/cm3

). The maximum activity value found for carbon-14 was reported in the analyses of 

soot from the flue gas collector (Wheelock, 1970 Appendix B, Table B-2), which is equal to 2.1E-02 

μCi/cm3

. The ratio of this activity to the total activity is 1.9(2.1E-02 μCi/cm3 ÷ 1.1E-02 μCi/cm3 = 

1.9), which would result in a process airborne concentration of 1.7E-08 μCi/cm3

 (1.9 X 9E-09 

μCi/cm3 = 1.7E-08 μCi/cm3). These concentrations can be used to assess the respective radionuclide 

exposures. The application of this method supports NIOSH’s ability to bound the internal dose from 

exposure to these radionuclides for the evaluated class. 

Table 7-1: Radionuclides of Concern by Category 

Coolant Impurities/ 

Components 

Corrosion 

Product Recoil Products 

Na-24 Mn-56 Na-24 

P-32 - Co-58 

S-35 - Co-60 

Cl-38 - - 

Mn-56 - - 
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As-76 - - 

Tritium - - 

C-14 - - 

The assessment methods presented here define the methods by which a maximum internal dose 

estimate can be determined for the evaluated worker class, which supports NIOSH’s conclusion that 

the operationally-related internal dose for the evaluated worker class can be bounded. 

7.2.3.2 Methods for Bounding Ambient Environmental Internal Dose

The method for bounding the ambient environmental internal dose is based on the conclusion that the 

total airborne activity in non-process areas would not have exceeded the lowest applicable nonprocess MPC without being detected by the existing monitoring system and noted in the routine 

summary reports. This conclusion is based on information in the SRDB and interviews with former 

POMR facility personnel, which indicated that the POMR facility had, and used, environmental and 

onsite sampling and analysis procedures with sufficient sensitivity to detect airborne particulate 

activity to the MPC levels in effect at the time (as noted in 10 C.F.R. pt. 20, Appendix B Table II). 

Typically, action levels for environmental monitoring are established at some percentage of the 

applicable limit for the most restrictive radionuclide (i.e., lowest MPC). For the purpose of bounding 

the ambient environmental dose, it will be assumed that the action level is equal to the MPC for the 

most restrictive radionuclide; the radionuclide with the lowest MPC listed in Table 5-2 is cobalt-60 

(3E-10 μCi/cm3 [10 CFR pt 20, 1961]). If the total beta/gamma airborne activity did not exceed 3E10 μCi/cm3

, and the total particulate activity consisted of 48% impurities, 13% corrosion products, 

and 39% recoil products (as noted in Section 5.2.1), the total air activity would be comprised of 1E10 μCi/cc impurities, 4E-11 μCi/cc corrosion products, and 1E-10 μCi/cc recoil products. 

The ratios calculated above (between the expected activity of tritium and carbon-14 and the total 

measured coolant activity) are used to determine ambient environmental airborne levels. For tritium, 

the expected airborne activity would be 5.7E-09 μCi/cm3

 (19.1 x 3E-10 μCi/cm3). For carbon-14, the 

expected ambient airborne level would be 5.7E-10 μCi/cm3

 (1.9 x 3E-10 μCi/cm3 = 5.7E-10 

μCi/cm3). 

As discussed in Section 7.2.2 and presented in this assessment of the ambient environmental airborne 

concentrations (as compared to the operationally-related values), the ambient environmental dose, 

based on the airborne activity levels outside the process areas and the plant grounds (and associated 

internal doses), can be bounded by the operationally-related values/levels and internal dose 

assessment methods. 

7.2.4 Internal Dose Reconstruction Feasibility Conclusion 

An extensive review of the data available on the SRDB, as well as interviews with former POMR 

facility personnel, have revealed no evidence of a radiological condition, transient or otherwise, that 

would lead to significant internal dose for the class evaluated in this report. In addition, information 

was located that leads to the conclusion that workplace monitoring, with the exception of tritium and 

carbon-14, was sufficiently sensitive to detect airborne radioactivity at levels lower than the MPC 

values in effect during the operational period for the facility. The tritium and carbon-14 

concentrations were evaluated through the determination of coolant concentration ratios, based on 
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what would be considered worst-case calculated conditions as defined in the Final Safeguards 

Summary Report for the Piqua Nuclear Power Facility (Atomics International, 1961). Based on this 

information and the assessment as presented in Section 7.2 of this report, NIOSH has concluded that it 

is feasible to bound the internal dose (reconstruct dose with sufficient accuracy) for the class 

evaluated in this report. 

7.3 Evaluation of Bounding External Radiation Doses at POMR 

ATTRIBUTION: Section 7.3 and its related subsections were completed by Louise Buker, Oak Ridge 

Associated Universities (ORAU) and Roger Halsey, Oak Ridge Associated Universities (ORAU). 

These conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for 

all conclusions in this document are explained in the associated text. 

The principal sources of external radiation doses for members of the proposed class were the Piqua 

Organic Moderated Reactor and the calibration sources used to calibrate the radiation monitoring 

equipment. 

The following subsections address the ability to bound external doses, methods for bounding doses, 

and the feasibility of external dose reconstruction. 

7.3.1 Evaluation of Bounding Process-Related External Doses 

The following subsections summarize the extent and limitations of information available for 

reconstructing the process-related external doses of members of the class under evaluation. 

7.3.1.1 Personnel Dosimetry Data

The available external dosimetry data for the site is in the form of summary data reported to the AEC 

Chicago Operations office for the City of Piqua workers and to the AEC Idaho Operations for the 

Atomics International workers who worked at the POMR facility. This summary data cover the 

complete timeframe and are based on film badge data. For the purpose of bounding the possible 

external doses, the external data appear to be reliable in summary form. The data are presented in 

Table 6-1 above. According to the references reviewed and an interview with a former Piqua health 

physics technician, the film badge exchange frequency was monthly for operation and maintenance 

personnel, and quarterly for administrative personnel (Personal Communication, 2009g). 

In addition to the summary data, there are three monthly reports that state the maximum exposure for 

the month, and a complete set of semiannual reports that make reference to employee external 

radiation exposures. Although the data are not fully comprehensive, whenever the maximum levels 

are mentioned, the maximum levels are much lower than the values in the AEC summary reports for 

the same time period. 

Photon
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The data are sufficient to support bounding photon dose estimates for the years 1963 through 1966 

from the upper bound of 1 rem for the years 1963 through 1965, and 2 rem for the year 1966; see 

Table 6-1 of this report. 

Beta

There is evidence that film badge beta monitoring was performed during the operational period, as 

indicated in the summary data. External dosimetry for plant workers were film badges provided by R. 

S. Landauer Jr. and Company (Atomics International, 1965b, p. 23), which included a shallow dose 

component. 

Documentation of shallow dosimetry results is incomplete for the operational lifetime of the POMR. 

In all of the documents that have been reviewed (City of Piqua, 1964), there is only one report of beta 

dose where gamma exposure is discussed (i.e., 60 mrem gamma and 20 mrem beta). 

Neutron

There is evidence that POMR facility personnel were monitored for neutron exposure with film 

badges (City of Piqua, 1964). During an interview, a former plant health physics technician said that 

he did not remember any positive neutron results from the Nuclear Track Emulsion (NTA), Type A 

film badges (Personal Communication, 2009g). Although there are no complete sets of neutron film 

badge results available, it is assumed that the complete set of summary data (reported to the AEC as 

whole-body radiation exposures from penetrating radiation) would have included neutron dose if it 

had been detected (AEC, 1963; AEC, 1964; AEC, 1965a; AEC, 1966). 

The limitation of using this penetrating external exposure summary data as bounding is due to the lack 

of sensitivity of NTA film dosimeters at lower energies (Fix, 1997); NTA film has a lower energy 

threshold (~500 keV, below which the measured dose rapidly decreases). The threshold of ~500 keV 

could cause the dosimeter to miss much of the neutron dose in highly shielded workplace radiation 

fields, including those in a reactor facility. The shielding significantly reduces the neutron energy, 

altering the exposure spectrum. In general, if the calibration spectra is the same as the workplace 

spectra then an accurate neutron dose can be measured (i.e., the same fraction of the neutron spectra is 

missed in the calibration and in the workplace). However, in highly shielded facilities, the workplace 

neutron spectra is much more degraded than the calibration spectra (Fix, 1997, pp. 39-45). 

Since the sensitivity of the NTA film is regarded as likely underestimating the neutron dose, the 

paired measurements of neutron and photon dose will be used to establish a neutron-to-photon ratio. 

Survey measurements reported in Piqua Nuclear Power Facility Operations Analysis Program 

Progress Report No. 4 (January 1, 1964-June 30, 1964) showed no detectable neutrons in normallyaccessible areas. In areas where neutrons were detected, the levels were low when compared to the 

potential for gamma exposure. Two measured values, each indicated as <0.5 mrem/hr, were found. 

These <0.5 mrem/hr readings were taken at the “OC-104, 14” line in the P-1B room location with a 

concurrent gamma reading of 13.5 mR/hr, and at the “OC-101, 14” line in the P-1A room location 

with a concurrent gamma reading 11 mR/hr (Atomics International, 1964a, p. 83). 
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The survey (reported in Piqua Nuclear Power Facility Operations Analysis Program Progress Report 

No. 4 (January 1, 1964-June 30, 1964 ) was conducted when the plant was operating at full power. 

The first detectable reading was associated with a gamma level of 13.5 mR/hr, assuming a worst case 

neutron reading of 0.5 mrem/hr, with a resulting neutron-to-photon ratio of 3.7% (or 0.037 to 1). The 

second detectable reading was associated with a gamma level of 11 mR/hr, with a worst case neutron 

reading of 0.5 mrem/hr, with a resulting neutron-to-photon ratio of 4.5% (or 0.045 to 1). Based on 

this information, a neutron-to-photon ratio of 10% (or 0.10 to 1) can be applied as a bounding ratio (in 

round numbers) to assess neutron dose given a photon dose (which would be assigned based on the 

bounding photon dose approach already discussed). Since these measurements were made with a long 

counter (Personal Communication, 2009g), which is known for being neutron energy 

independent (Fix, 1997; Cember, 1969), this approach avoids any energy dependence issues related to 

the NTA badges. 

7.3.1.2 Area Monitoring Data

Area monitoring data were available from progress reports, monthly reports, and semiannual reports. 

Photon

In the monthly, semiannual, and annual reports, there are references to ambient gamma levels, 

primarily presented as example levels. For example, Piqua Nuclear Power Facility Operations 

Analysis Program Progress Report No. 4 (January 1, 1964-June 30, 1964) indicated that “Gamma 

radiation which penetrated the biological shield was less than 1 mr/hr” (Atomics International, 1964a, 

p. 78). However, there are no consistent surveys or summary data available. 

The most complete set of monitoring data available is in a table summarizing a survey taken in 1964 

(during the first year of operation), described as a “typical survey.” The table includes readings from 

eighteen locations that ranged from 0.1 mr/hr at the “FELS Tank” location to 250 mr/hr at the “F-2A 

Degasifier filter” location (Atomics International, 1964a, p. 83). The report indicated that with the 

exception of the degasifier filter, radiation levels remained low enough to permit operator access for 

contact maintenance on the primary coolant system without time limits (Atomics International, 1964a, 

p. 82). 

Beta

There are several discrete cases mentioned in the reviewed documents where beta-gamma radiation 

levels were measured (Atomics International, 1967, p. 15; Atomics International, 1965c, p. 76; 

Atomics International, 1964a, p. 17). NIOSH discovered only a couple of instances where the betagamma reading was distinguished from the gamma activity (Atomics International, 1964a, p. 17; 

Atomics International, 1964c, pp. 19-20). 
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Neutron

In an interview with a former POMR plant health physicist technician, the former employee stated that 

neutron surveys were routinely performed and were performed at every power level, and that the type 

of detector used at the plant was a long counter (Personal Communication, 2009a; Personal 

Communication, 2009g) 

One of these surveys, conducted while the plant was operating at full power, was reported in Piqua 

Nuclear Power Facility Operations Analysis Program Progress Report No. 4 (January 1, 1964-June 

30, 1964); this report included a summary table listing the results of a “typical radiation survey” of the 

plant when it was operating at 45.4 MWt (Atomics International, 1964a). The neutron radiation 

surveys of the “Reactor 100-ft level” location, the “East Reactor Biological Shield at the 78’ level” 

location, and the “East Reactor Shield at the 72’ level” location were indicated as having “no 

neutrons” and were recorded as < 0.5 mrem/hr neutrons (Atomics International, 1964a). 

Although neither the instrument nor the methodology is mentioned, an inference may be made about 

the detection level by the two non-zero values. The minimum detection level of the process would be 

a value greater than zero, but at most, 0.5 mrem per hour. 

7.3.2 Evaluation of Bounding Ambient Environmental External Doses 

The ambient environmental external dose is accounted for, and therefore bounded, by the operational 

dose. Therefore, further discussion and analysis of the external ambient environmental dose will not 

be included in this report. 

7.3.3 POMR Occupational X-Ray Examinations 

NIOSH has found no records indicating that employees at Piqua Organic Moderated Reactor Facility

were required to complete medical examinations, including chest X-rays prior to beginning work, on a 

periodic basis (e.g., annually), or following termination. However, a former Piqua shift supervisor 

indicated that chest X-rays were included in the annual physical (Personal Communication, 2009c). 

Although no records have been identified that indicate that occupational medical X-rays were 

required, the dose associated with X-ray exams can be assessed using the methodology defined in 

ORAUT-OTIB-0006. NIOSH believes that this methodology supports its ability to bound the 

occupational medical X-ray doses for the evaluated class. 
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7.3.4 Methods for Bounding External Dose at POMR 

There is an established protocol for assessing external exposure when performing dose reconstructions 

(these protocol steps are discussed in the following subsections): 

• Photon Dose 

• Beta Dose 

• Neutron Dose 

• Medical X-ray Dose 

7.3.4.1 Methods for Bounding Operational Period External Dose

Photon Dose

The photon dose can be assessed based on bounding assumptions from the site summary data for 

external dose (based on information provided in Table 6-1, above). The photon information, as 

discussed in the personal dosimetry and area monitoring data sections above, supports applying the 

bounding dose of 1 rem for the years 1963 through 1965 and 2 rem for the year 1966. The upper 

bound for a particular year would be assigned to individuals who worked at the POMR site. 

Beta Dose

Survey data reported in the annual and semiannual reports were typically reported as beta-gamma 

(i.e., β-γ). There were a couple of instances where the gamma exposure rate was distinguished from 

the beta-gamma total. These instances described scenarios with the potential for personnel exposure. 

The first instance described a measurement on an in-core filter that had been removed. It had a 

reading of 20 mrad/hr beta-gamma and 5 mR/hr gamma (Atomics International, 1964a, p. 17). The 

second instance was measured as 10 mrads/hr beta-gamma, including 0.5 mR/hr gamma at 1 

inch (Atomics International, 1964c, pp. 19-20). 

The Varskin 3 software application was also used to assess bounding beta-to-gamma ratios for direct 

contact (for hands-on) dose to the extremities, based on the evaluation of source term information. 

Gamma spectroscopy results from three coolant filter media were used as the basis of the analysis; 

cobalt-58, cobalt -60, iron-59, manganese-54, and zinc-65 were identified in the gamma spectroscopy 

results (Atomics International, 1964a, p. 102). The Varskin 3 analysis determined a ratio of 40:1 for 

direct contact with the skin (0 cm air gap) and a 20:1 ratio for a 5 cm air gap (the largest gap allowed 

by Varskin 3). A bounding beta-to-gamma ratio of 40:1 can be applied when calculating direct 

contact dose to the hand. This is a bounding value because this assumes direct material contact (i.e., 

coolant) and that no personal protective equipment (e.g., gloves) was worn. Furthermore, a bounding 

beta-to-gamma ratio of 20:1 can be applied to the upper extremities (e.g., elbow to hand) or lower 

extremities (e.g., legs, below the knee), in cases of direct contact. The beta-to-gamma ratio of 20:1 for 

the extremities (excluding the hand) is considered bounding because the material would need to be 

within 5 cm of the body. Furthermore, the extra shielding from personal protective equipment is not 

taken into account; thus, the potential beta fractions would be overestimated for the shallow dose 

calculation. 
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Neutron Dose

The potential for neutron dose existed only when the reactor was operating. The potential for neutron 

exposure is not independent of the potential for gamma exposure; that is, there was no place around 

the reactor where personnel could be exposed to neutrons that didn’t also have gamma exposure. 

Measurements reported in Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 4 (January 1, 1964-June 30, 1964) showed no detectable neutrons in normally-accessible 

areas. In areas where neutrons were detected, the levels were low when compared to the potential for 

gamma exposure. Two values, each indicated as <0.5 mrem/hr, were found. These <0.5 mrem/hr 

readings were taken at the “OC-104, 14” line in the P-1B room” location with a concurrent gamma 

reading of 13.5 mR/hr and at the “OC-101, 14” line in the P-1A room” location with a concurrent 

gamma reading of 11 mR/hr (Atomics International, 1964a, p. 83). 

This survey reported in Piqua Nuclear Power Facility Operations Analysis Program Progress Report 

No. 4 (January 1, 1964-June 30, 1964) was conducted when the plant was operating at full power, and 

should represent the maximum neutron flux. The first detectable reading was associated with a 

gamma level of 13.5 mR/hr, giving a neutron-to-photon ratio of 3.7% (or 0.037 to 1). The second was 

associated with a gamma level of 11 mR/hr, giving a neutron-to-photon ratio of 4.5% (or 0.045 to 1). 

Based on this information, a neutron-to-photon ration of 10% (or 0.10 to 1) can be applied as a 

bounding ratio to assess neutron dose given a photon dose (which would be assigned based the 

bounding photon dose approach section above). 

This value is conservative as (1) the actual reported value was “<0.5 mrem/hr”, (2) the survey showed 

no measureable results in the normally occupied areas of the plant, (3) the plant was operating at full 

power, and (4) the ratio is more than twice the value of the measured ratios. In addition, this is 

corroborated by a statement in an interview with a former plant health physics technician who stated 

that he did not recall a positive neutron result on any film badge report and that the only time he 

observed detectable neutron exposure was when calibrating the neutron survey instrument. (Personal 

Communication, 2009g). 

7.3.5 External Dose Reconstruction Feasibility Conclusion 

Although no personal dosimetry data were available, NIOSH has access to sufficient information to 

support assessing the external dose for the class evaluated in this report. The method for bounding the 

operational period external dose is based on bounding the doses with the upper limit of the summary 

data for photon dose and through the application of bounding beta-to-gamma and neutron-to-photon 

ratios, which were based on surveys that represent maximum exposure scenarios and source term 

information. Based on this information and the assessment as presented in Section 7.3, NIOSH has 

concluded that it is feasible to bound the external dose (reconstruct dose with sufficient accuracy) for 

the class evaluated in this report. 
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7.4 Evaluation of Petition Basis for SEC-00126 

ATTRIBUTION: Section 7.4 and its related subsections were completed by Karin Jessen, Oak Ridge 

Associated Universities (ORAU). These conclusions were peer-reviewed by the individuals listed on 

the cover page. The rationales for all conclusions in this document are explained in the associated 

text. 

The following subsections evaluate the assertions made on behalf of petition SEC-00126 for the 

POMR site. 

7.4.1 Covered Period of Employment 

ISSUE: The petitioner stated that he had received a letter from DOL indicating to him that the covered 

period of employment would be expanded through 1969.

RESPONSE: NIOSH was unable to confirm any letter from DOL about expanding the covered period. 

However, NIOSH submitted a letter to DOL requesting that the covered period include the D&D 

period. As of April 13, 2009, NIOSH has not received any additional DOL direction regarding a 

change in the covered employment period for the POMR facility. 

7.4.2 Monitoring 

 

ISSUE: ...He was a laborer primarily assigned any duty his supervisor would assign. This all done 

without any type of monitoring, training, or protective devices for handling nuclear material this was 

not explained but one has to surmise it was because of [Name Redacted] primary work was at the 

other building. 

RESPONSE: Although NIOSH has not located individual data, NIOSH is aware that radiological 

monitoring was performed at the POMR facility. This is evident in the monthly and semiannual 

reports where exposures were reported and summarized in the reports. In addition, the people 

interviewed for this report stated that they remember wearing monitoring devices, including pencil 

dosimeters and film badges (Personal Communication, 2009a; Personal Communication, 2009c;

Personal Communication, 2009d; Personal Communication, 2009e). For any potential unmonitored 

Piqua personnel, it is assumed that exposures calculated or measured for monitored Piqua personnel 

would exceed any potential exposures to unmonitored personnel. 

7.5 Summary of Feasibility Findings for Petition SEC-00126 

This report evaluates the feasibility for completing dose reconstructions for employees at the POMR 

facility from January 1, 1963 through May 1, 1966. NIOSH found that the available summary 

records, process descriptions, and source term data available are sufficient to complete dose 

reconstructions for the evaluated class of employees. 

Table 7-2 summarizes the results of the feasibility findings at POMR for each exposure source during 

the time period from January 1, 1963 through May 1, 1966. 
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Table 7-2: Summary of Feasibility Findings for SEC-00126 

January 1, 1963 through May 1, 1966 

Source of Exposure Reconstruction Feasible Reconstruction Not Feasible 

Internal X 

 Process Related X 

 Ambient/ Environmental X 

External X 

 - Gamma X 

 - Beta X 

 - Neutron X 

 - Occupational Medical X-ray X 

8.0 Evaluation of Health Endangerment for Petition SEC-00126 

The health endangerment determination for the class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(3). Under these requirements, if it is not 

feasible to estimate with sufficient accuracy radiation doses for members of the class, NIOSH must 

also determine that there is a reasonable likelihood that such radiation doses may have endangered the 

health of members of the class. Section 83.13 requires NIOSH to assume that any duration of 

unprotected exposure may have endangered the health of members of a class when it has been 

established that the class may have been exposed to radiation during a discrete incident likely to have 

involved levels of exposure similarly high to those occurring during nuclear criticality incidents. If 

the occurrence of such an exceptionally high-level exposure has not been established, then NIOSH is 

required to specify that health was endangered for those workers who were employed for a number of 

work days aggregating at least 250 work days within the parameters established for the class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC. 

Based on POMR process information, summary report information, source term information, and 

interviews with former POMR facility personnel, NIOSH’s evaluation determined that it is feasible to 

estimate radiation dose for members of the NIOSH-evaluated class with sufficient accuracy based on 

the sum of information available from available resources. Modification of the class definition 

regarding health endangerment and minimum required employment periods, therefore, is not required. 

9.0 Class Conclusion for Petition SEC-00126 

Based on its full research of the class under evaluation, NIOSH found no part of said class for which it 

cannot estimate radiation doses with sufficient accuracy. This class includes all employees associated 

with reactor activities who worked within and around the reactor dome at the Piqua Organic 

Moderated Reactor during the covered period from January 1, 1963 through May 1, 1966. 
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NIOSH has carefully reviewed all material sent in by the petitioner, including the specific assertions 

stated in the petition, and has responded herein (see Section 7.4). NIOSH has also reviewed available 

technical resources and many other references, including the Site Research Database (SRDB), for 

information relevant to SEC-00126. In addition, NIOSH reviewed its NOCTS dose reconstruction 

database to identify EEOICPA-related dose reconstructions that might provide information relevant to 

the petition evaluation. 

These actions are based on existing, approved NIOSH processes used in dose reconstruction for 

claims under EEOICPA. NIOSH’s guiding principle in conducting these dose reconstructions is to 

ensure that the assumptions used are fair, consistent, and well-grounded in the best available science. 

Simultaneously, uncertainties in the science and data must be handled to the advantage, rather than to 

the detriment, of the petitioners. When adequate personal dose monitoring information is not 

available, or is very limited, NIOSH may use the highest reasonably possible radiation dose, based on 

reliable science, documented experience, and relevant data to determine the feasibility of 

reconstructing the dose of an SEC petition class. NIOSH contends that it has complied with these 

standards of performance in determining the feasibility or infeasibility of reconstructing dose for the 

class under evaluation. 
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10.0 References 

10 C.F.R. pt. 20, Standards for Protection Against Radiation, Cumulative Supplement of the Federal 

Register; January 1, 1961; SRDB Ref ID: 62588 

42 C.F.R. pt. 81, Guidelines for Determining the Probability of Causation Under the Energy 

Employees Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol. 67, No. 85/Thursday, p 22,296; May 2, 2002; SRDB Ref ID: 19391 

42 C.F.R. pt. 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule; May 2, 2002; SRDB Ref ID: 

19392 

42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as Members of the Special 

Exposure Cohort Under the Energy Employees Occupational Illness Compensation Program Act of 

2000; Final Rule; May 28, 2004; SRDB Ref ID: 22001 

42 U.S.C. §§ 7384-7385 [EEOICPA], Energy Employees Occupational Illness Compensation 

Program Act of 2000, as amended 

OCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 0; 

National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; September 23, 

2004; SRDB Ref ID: 32022 

ORAUT-OTIB-0006, Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures; Oak Ridge Associated Universities (ORAU); Oak Ridge, Tennessee; December 21, 2005; 

SRDB Ref ID: 20220 

AEC, 1963, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1963, results for Piqua; Atomic Energy Commission (AEC); 1963; 

SRDB Ref ID: 13792, p. 34 

AEC, 1964, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1964, results for Piqua; Atomic Energy Commission (AEC); 1964; 

SRDB Ref ID: 14442, p. 18 

AEC, 1965a, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1965, results for Piqua; Atomic Energy Commission (AEC); 1965; 

SRDB Ref ID: 13797, p. 27 

AEC, 1965b, Radiation Exposure History for One Employee, 1961-1965; U.S. Atomic Energy 

Commission (AEC); 1965; SRDB Ref ID: 62522, p. 3 

AEC, 1966, Summary of Whole Body Radiation Exposures to External Penetrating Radiation 

Accumulated During the Year-1966, results for Piqua; Atomic Energy Commission (AEC); 1966; 

SRDB Ref ID: 13798, p. 23 
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Arnold, 1966, Irradiation Performance of Piqua Core 1 Fuel Elements, NAA-SR-11712; J. L. 

Arnold; May 25, 1966; SRDB Ref ID: 53719 

Ashley, 1964, Reactor Information, correspondence to J. Jacobson; R. L. Ashley; January 10, 1964; 

SRDB Ref ID: 57974, p. 173 

Atomics International, unknown date, Piqua Nuclear Power Facility Progress Report from Criticality 

to First Scheduled Shutdown (June 1963-May 1964); Atomics International; unknown date; SRDB 

Ref ID: 56012 

Atomics International, 1961, Final Safeguards Summary Report for the Piqua Nuclear Power 

Facility, NAA-SR-5608 Reactor Technology; Atomics International; August 1, 1961; SRDB Ref ID: 

53945, pp. 12-331 

Atomics International, 1963, Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 2 (Fiscal Year 1963), NAA-SR-8722 Reactor Technology; Atomics International; 

November 30, 1963; SRDB Ref ID: 55926 

Atomics International, 1964a, Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 4 (January 1, 1964-June 30, 1964), NAA-SR-10307 Reactor Technology; Atomics 

International; December 15, 1964; SRDB Ref ID: 55814 

Atomics International, 1964b, Piqua Nuclear Power Facility Report of Power Operation; Atomics 

International; March 6, 1964; SRDB Ref ID: 56010 

Atomics International, 1964c, Piqua Nuclear Power Facility Monthly Operating Report No. 19; 

Atomics International; November 1964; SRDB Ref ID: 55945 

Atomics International, 1964d, Piqua Nuclear Power Facility Operations Analysis Program Progress 

Report No. 3 (July 1, 1963-December 31, 1963), NAA-SR-9473 Reactor Technology; Atomics 

International; July 15, 1964; SRDB Ref ID: 55932 

Atomics International, 1965a, Compilation of Piqua Nuclear Power Facility Operating Limits and 

Controls and Post-Critical Operational Tests; Atomics International; February 5, 1965; SRDB Ref 

ID: 55946 

Atomics International, 1965b, Piqua Nuclear Power Facility Monthly Operating Report No. 24; 

Atomics International; April 1965; SRDB Ref ID: 55943 

Atomics International, 1965c, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 6 (January 1-June 30, 1965); Atomics International; December 15, 

1965; SRDB Ref ID: 55817 

Atomics International, 1965d, Piqua Nuclear Power Facility Monthly Operating Report No. 26; 

Atomics International; June 1965; SRDB Ref ID: 55944 
SEC-00126 04-27-09 Piqua Organic Moderated Reactor (POMR)

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Atomics International, 1965e, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 5 (July 1-December 31, 1964); Atomics International; June 15, 

1965; SRDB Ref ID: 55815 

Atomics International, 1966a, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 7 (July 1, 1965-January 13, 1966); Atomics International; October 

25, 1966; SRDB Ref ID: 55880 

Atomics International, 1966b, Piqua Nuclear Power Facility Reactor Operations Analysis Program 

Semiannual Progress Report No. 8 (January 1-June 30, 1966), NAA-SR-12148 General, 

Miscellaneous; Atomics International; December 31, 1966; SRDB Ref ID: 55881 

Atomics International, 1967, Piqua Nuclear Power Facility Monthly Operating Report No. 53, COO652-43; Atomics International; September 1967; SRDB Ref ID: 55942 

Cember, 1969, Introduction to Health Physics, 1st edition; Herman Cember; January 1969; SRDB Ref 

ID: Publicly available 

City of Piqua, 1964, 1963 Radiation Exposure Records for AI Personnel at the PNPF, 

correspondence; City of Piqua; February 11, 1964; SRDB Ref ID: 62522, p. 2 

DOE, 2009, Piqua Nuclear Power Facility; DOE Environmental Management website; last accessed 

on March 20, 2009; SRDB Ref ID: 53525 

Fix, 1997, Retrospective Assessment of Personnel Neutron Dosimetry for Workers at the Hanford Site, 

PNNL-11196 UC-606; J. J. Fix, R. H. Wilson, and W. V. Baumgartner; February 1997; SRDB Ref 

ID: 5275 

Form B, 2008, Petition Form B; August 21, 2008; OSA Ref ID: 106811 

Geiger, 1969, Piqua Nuclear Power Facility Final Survey Report; Eric L. Geiger; January 23, 1969; 

SRDB Ref ID: 54900, pp. 18-25 

Hewson, 1969, PNPF Retirement Safety Analysis Reevaluation of Residual Nuclides, AI-AECMEMO-12708 Supplement A; R. A. Hewson; January 15, 1969; SRDB Ref ID: 58051 

McCurnin, 1966, Visual Examination of Fuel Elements at PNPF through May 19, 1966, 

correspondence to F. C. Grouemeyer; W. R. McCurnin; SRDB Ref ID: 57303, pp. 25-54 

Morgan, 1966, Piqua Recovery Program—Phase II-A Plan of Action; W. T. Morgan; July 11, 1966; 

SRDB Ref ID: 57303, pp. 104-134 

Personal Communication, 2009a, Personal Communication with Former Health Physics Technician; 

Telephone Interview by ORAU Team; January 20, 2009; SRDB Ref ID: 61683 
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Personal Communication, 2009b, Personal Communication with DOE Legacy Management 

Employee; Telephone Interview by ORAU Team; January 29, 2009; SRDB Ref ID: 61681

Personal Communication, 2009c, Personal Communication with Former Shift Supervisor; Telephone 

Interview by ORAU Team; February 19, 2009; SRDB Ref ID: 61677

Personal Communication, 2009d, Personal Communication with Former Construction Engineer, 

Instrumentation Engineer, and Electrical Engineer; Telephone Interview by ORAU Team; February 

19, 2009; SRDB Ref ID: 61684 

Personal Communication, 2009e, Personal Communication with Former Reactor Operator and 

Maintenance Foreman; Telephone Interview by ORAU Team; February 6, 2009; SRDB Ref ID: 

61679

Personal Communication, 2009f, Personal Communication with Chief Health Physicist; Telephone 

Interview by ORAU Team; February 23, 2009; SRDB Ref ID: 61680 

Personal Communication, 2009g, Personal Communication with Former Health Physics Technician, 

second interview; Telephone Interview by ORAU Team; March 18, 2009; SRDB Ref ID: 62597

Personal Communication, 2009h, Personal Communication with Former POMR Facility Employee; 

documented telephone communication; March 23, 2009; SRDB Ref ID: 62596

Radiological Health Handbook, 1970, Radiological Health Handbook; U.S. Department of Health, 

Education, and Welfare, compiled and edited by the Bureau of Radiological Health and the Training 

Institute, Environmental Control Administration; January 1970; SRDB Ref ID: publicly available 

Schwartz, 1983, Experimental Verification of the Neutron Spectrum from the NBS D/Sub 2/OModerated/Sup 252/Cf Source, NUREC/CR-3399; R. B. Schwartz, C. M. Eisenhauer, and J. A. 

Grundl; August 1, 1983; SRDB Ref ID: 64008 

Unknown author, unknown date-a, Summary of Operating History, select pages from another 

unidentified document; unknown date; SRDB Ref ID: 21882 

Unknown author, unknown date-b, Piqua Nuclear Power Facility History; unknown author; unknown 

date; SRDB Ref ID: 53751 

Wheelock, 1970, Retirement of the Piqua Nuclear Power Facility, AI-AEC-12832 Reactor 

Technology; C. W. Wheelock; April 1, 1970; SRDB Ref ID: 21860 
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Attachment One: Information from Interviews with Former Piqua 

Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

Shift Supervisor, City of Piqua, July 1960-February 1969 (PC, 2009c) 

Pencil dosimeters were used during normal operations (Stored in rack when not in use). Film badges were used when fuel came on the site to load the core (until the fuel was put in storage in 1969). 

Routinely No neutrons to be 

concerned about. 

Beta-gamma was the only concern. 

CAMS were not used; the only air monitoring done was in the ventilation system. 

Moderator temperature kept increasing during power operation. [Note: Interviewee thought this was kept hidden, but NIOSH found an entire write up dedicated to this.]

Possibly done occasionally. 

HP Technician, City of Piqua, Fall 1961-Spring 1966 (PC, 2009a & PC, 2009g) 

When operations began, all workers wore film badges that were exchanged monthly. Admin workers badges were eventually exchanged quarterly. Of the ~45 workers who were routinely monitored, only 1 or 2 workers would have a positive reading (20-30 mR) above MDL. Control badges were used to determine background levels. Thinks NTA badges used for neutron exposures. Visitors were assigned film badges. As many as 31 workers from Atomics International, Inc. (AI) worked at Piqua because they were responsible for reactor startup. Some of 

Daily, weekly, and monthly surveys, depending on type of survey. All surveys were documented. Neutron surveys were done routinely, at least monthly. When the power level was raised, extensive neutron surveys were done initially. Afterwards, the surveys were less extensive. 

No neutron issues; the greatest neutron dose came from calibrating the neutron survey instrument using a PuBe source. There was a potential for neutron exposure at the primary coolant line (P1B). The levels were less than 0.5 mrem/hr. 

They could do all maintenance work at any time because there was not a dose issue. The only measurable exposure occurred when they were working with the instrument thimbles on top of the reactor in the high-bay area. When tubular fuel rods were pulled, they were about 200 mR/hr. HPs left them on ground, took a coffee break to allow for radioactive decay. Filter change out comment: It was a hot (thermal) and dirty (nonradioactive) job, but the exposure rate was low. The highest gamma dose rate (10-15 mR/h) 

CAMs operating at all times. Rotating filter media was changed hourly and was read continuously using an end window GM tube. It was all recorded on a strip chart recorder. One sample was taken every month and sent away for outside analysis. A portion of the filter was sent out monthly for analysis and then sent to AEC Chicago. 

No unusual events that led to either contamination or exposure. 

Does not recall abioassays or WBbeing performed. 
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Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

these workers wore both AI and Piqua film badges, but Piqua was responsible for tracking their doses. Dosimetry records first went to the City of Piqua, and then were sent to AEC Chicago Operations Office on a monthly basis. All badges were shipped to Landauer in Chicago. 

was around the primary coolant pump coming out of the core. The gamma dose was very small and there was usually nothing to report. No worker had a monthly dose exceeding 50 mR. Cannot remember the highest annual dose, but it was so minimal that it was not a concern. 

Construction Engineer, Atomics International, 1961-1966 (PC, 2009d) 

All reactor operators and maintenance workers wore film badges all the time, and they also had pencil dosimeters. Recalled that administrative workers had to go to an HP to get a film badge if they were going into the Reactor Building. Years after he left, he received information from the AEC that told him how much radiation dose he received. He said that the AEC would send exposure records to AI. 

Contamination surveys were routinely performed; it was a very clean plant. HPs would tell workers where rad levels were. 

Neutrons in core, but no leaks of neutrons from reactor vessel. Containment tight and 6’ of shielding around reactor. 

Was a very, very low-radiation-level place. 

Does not recall any area sampling being performed, but the exhaust air from the stack was monitored. 

Around 1964, in the Reactor Building, a seal on a pump failed and began spewing out organic material coolant. When the seal let loose and the coolant spilled out, supposedly there was a fire, but no one saw it. 

-

Reactor Operator & Maintenance 

Does not recall film badge exchange frequency; it may have 

Radiation and contamination surveys were 

No neutron radiation. 

The beta and gamma radiation were from 

Air monitoring was done inside and 

-

Does not recall giving urine samples. Howev
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Dosimetry Survey Neutron General Radiation CAMs Incidents Bioassay 

Foreman, DOE, 1960-1969 (PC, 2009e) 

been done monthly. Never saw the results of film badge measurements, but he received a yearly letter telling him what his exposure was for that year. Does not recall getting any exposure. The pencil dosimeters were used for a whole day and registered a maximum reading of 200 mR. The workers turned the dosimeters over to the HP every day and the HP checked the reading, but he does not know if the readings were recorded. Does not recall ever getting a reading on his dosimeter. Although, he might have gotten a reading on one job when he cleaned up the residue on the absolute filters from the coolant on a big boiler. 

routinely performed. 

activation of the impurities in the coolant. 

outside the POMR plant. 

he was not saying did not happen, juthat he does not remember giving urine samples. 

Health Physicist, City of Piqua (PC, 2009f) 

- Only smear 

surveys of the floors. 

No neutron exposures. 

The highest doses at the site were about 100 mrem in a month. Doses could have also been in the 30-40 mrem/month range. 

No air contamination. - - 
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Attachment Two: Data Capture Synopsis 

Table A2-1: Data Capture Synopsis for POMR 

Data Capture Information Data Capture Description C

Primary Site/Company Name: Piqua Organic Moderated Reactor; DOE 1963-1966 Other company names:

Piqua Nuclear Power Facility PNPF Atomics International, now Boeing at Santa Field Laboratory, 1963-1964, see below for data capture results 

See below. See 

State Contacted: Ohio Department of Health, David Lipp 

No relevant documents not already in the SRDB identified. 11/0

American Municipal Power-Ohio, Kent Carson, Sr. 

No relevant documents identified. 02/0

Cincinnati Public Library An operational report from the first criticality to the first scheduled 

shutdown. 

11/0

Claimant Documented communications, monthly report #69, and a final survey 

report. 

02/2

Comprehensive Epidemiologic Data Resource (CEDR) 

No relevant documents identified. 10/1

Department of Labor (Paragon) Testing and operation reports at 50% and 100% power, core unloading and 

inspection report and drawings, facility photos, failure analyses, and 

cathodic protection of the containment shell. 

12/0

DOE Hanford Declassified Document Retrieval System (DDRS) 

No relevant documents identified. 10/1

DOE Legacy Management Considered Sites All Considered Sites documents are contained in the Grand Junction Office 

DVD's. No search was necessary. 

N/A 

DOE Legacy Management - Grand Junction Office 

Annual site radiological surveys, FUSRAP documents, total residual radioactivity in the reactor, building drawings, and remediation planning documents. 

11/2

DOE OpenNet No relevant documents identified. 10/1DOE OSTI Energy Citations Reports on shipping casks and waste treatment, and an Energy Research 

and Development Administration Decontamination and Decommissioning conference. 

01/1
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Table A2-1: Data Capture Synopsis for POMR 

Data Capture Information Data Capture Description C

DOE OSTI Information Bridge A 1991 annual site inspection and radiological survey, reports on the 

performance and metallurgy of organic moderated reactor fuel elements, and a study of military applications. 

01/1

Google FUSRAP site fact sheet, a discussion of entombment as a decommissioning 

technology, AEC and DOE reports, and a 2008 annual site inspection and radiological survey. 

10/2

Interlibrary Loan Facility engineering design and a status report. 12/1National Academies Press (NAP) No relevant documents identified. 10/1National Nuclear Security Administration (NNSA) - Nevada Site Office 

No relevant documents identified. 10/1

NRC Agencywide Document Access and Management (ADAMS) 

No relevant documents identified. 10/1

Office of Scientific & Technical Information (OSTI) 

Operations analysis progress reports, engineering studies of organic moderation, and information on preparation and shipping of fuel elements. 

12/1

Piqua Public Library No relevant documents identified. 01/2Science Applications International Corporation (SAIC) 

Complex-wide exposure records for 1960 and 1961. 09/0

San Bruno Federal Records Center Quality assurance documentation and weekly reports, all relating to the deactivation of the facility. 02/0Santa Susanna Field Laboratory (Boeing) Fuel element safety analysis, fuel shipment reports, criticality controls, 

radiological surveillance program, recovery plans, weekly highlights reports, studies of organic contaminant formation, 1967 recovery program plans, coolant surveillance reports, reactor operations analysis program reports, control rod studies, modification studies, coolant chemistry reports, coolant filtration studies, core examination reports, and facility retirement reports. 

12/0

Unknown Personnel exposure information for 1963 and film badge records for 2 

employees. 

07/3

Washington State University (U.S. Transuranium and Uranium Registries) 

No relevant documents identified. 10/1

Total 
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Table A2-2: Database Searches for POMR 

Database/Source Keywords 

DOE OpenNet http://www.osti.gov/opennet/advancedsearch.jsp COMPLETED 10/10/2008 

"Piqua" + 01/01/1960 - 08/10/2008 

DOE Hanford DDRS http://www2.hanford.gov/declass/ COMPLETED 10/10/2008 

"Piqua" + 01/01/1960 - 08/10/2008 

DOE OSTI Energy Citations http://www.osti.gov/energycitations/ COMPLETED 01/14/2009 

"Piqua" + Reactor + 01/01/1960 - 08/10/2008 

DOE OSTI Information Bridge http://www.osti.gov/bridge/advancedsearch.jsp COMPLETED 01/14/2009 

"Piqua" + 01/01/1960 - 08/05/2008 

DOE CEDR http://cedr.lbl.gov/ COMPLETED 10/10/2008 

"Piqua" 

NNSA - Nevada Site Office www.nv.doe.gov/main/search.htm COMPLETED 10/10/2008 

"Piqua" 

U.S. Transuranium & Uranium Registries http://www.ustur.wsu.edu/ COMPLETED 10/10/2008 

"Piqua" 

National Academies Press http://www.nap.edu/ COMPLETED 10/10/2008 

"Piqua" 

NRC ADAMS Reading Room http://www.nrc.gov/reading-rm/adams/webbased.html COMPLETED 10/10/2008 

"Piqua" + 01/01/1960 - 08/10/2008 

Google http://www.google.com COMPLETED 10/25/2008 

americium OR Am241 OR Am-241 OR "AM 241" OR 241Am OR 241-Am OR "241 Am" AND "Piqua" ionium OR Th230 OR Th-230 OR "Th 230" OR 230Th OR 230-Th OR "230 Th" AND "Piqua" neptunium OR Np237 OR Np-237 OR "Np 237" OR 237Np OR 237-Np OR "237 Np" AND "Piqua" 
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Table A2-2: Database Searches for POMR 

Database/Source Keywords 

polonium OR Po210 OR Po-210 OR "Po 210" OR 210Po OR 210-Po OR "210 Po" AND "Piqua" thorium OR Th232 OR Th-232 OR "Th 232" OR 232Th OR 232-Th AND "Piqua" AND Reactor "232 Th" OR "Z metal" OR myrnalloy OR "chemical 10-66" OR "chemical 10-12" AND "Piqua" ionium OR UX1 OR UX2 OR Th-230 OR Th230 OR "Th 230" OR 230-Th OR "230 Th" AND "Piqua" 230Th OR Th-234 OR Th234 OR "Th 234" OR 234-Th OR 234Th OR "234 Th" AND "Piqua" tritium OR H3 OR H-3 OR mint OR HTO AND "Piqua" AND Reactor uranium OR U233 OR U-233 OR "U 233" OR 233U OR 233-U OR "233 U" AND "Piqua" AND Reactor U234 OR "U 234" OR U-234 OR 234U OR 234-U OR "234 U" AND "Piqua" U235 OR "U 235" OR U-235 OR 235-U OR 235U OR "235 U" OR U238 AND "Piqua" "U 238" OR U-238 OR 238-U OR 238U OR "238 U" AND "Piqua" U308 OR "U 308" OR U-308 OR 308-U OR 308U OR "308 U" OR "uranium extraction" OR "black oxide" OR "brown oxide" AND "Piqua" "green salt" OR "orange oxide" OR "yellow cake" OR UO2 OR UO3 AND "Piqua" UF4 OR UF6 OR C-216 OR C-616 OR C-65 OR C-211 OR U3O8 AND "Piqua" AND Reactor 
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Table A2-2: Database Searches for POMR 

Database/Source Keywords 

plutonium OR Pu-238 OR Pu238 OR "Pu 238" OR 238Pu OR 238-Pu OR "238 Pu" AND "Piqua" AND Reactor Pu-239 OR Pu239 OR "Pu 239" OR 239Pu OR 239-Pu OR "239 Pu" AND "Piqua" Pu-240 OR Pu240 OR "Pu 240" OR 240Pu OR 240-Pu OR "240 Pu" AND "Piqua" Pu-241 OR Pu241 OR "Pu 241" OR 241Pu OR 241-Pu OR "241 Pu" AND "Piqua" radium OR Ra-226 OR Ra226 OR "Ra 226" OR 226-Ra OR 226Ra OR 226-Ra AND "Piqua" AND Reactor Ra-228 OR Ra228 OR "Ra 228" OR 228Ra OR 228-Ra OR "228 Ra" AND "Piqua" radon OR Rn-222 OR Rn222 OR "Rn 222" OR 222Rn OR 222-Rn OR "222 Rn" AND "Piqua" AND Reactor thoron OR Rn-220 OR Rn220 OR "Rn 220" OR 220Rn OR 220-Rn OR "220 Rn" AND "Piqua" AND Reactor protactinium OR Pa-234m OR Pa234m OR "Pa 234m" OR 234mPa OR 234m-Pa OR "234m Pa" AND "Piqua" AND Reactor strontium OR Sr-90 OR Sr90 OR "Sr 90" OR 90-Sr OR 90Sr OR "90 Sr" AND "Piqua" AND Reactor oralloy OR postum OR tuballoy OR "uranyl nitrate hexahydrate" OR UNH OR K-65 OR "sump cake" AND "Piqua" AND Reactor "uranium dioxide" OR "uranium tetrafluoride" OR "uranium trioxide" AND "Piqua" 
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Table A2-2: Database Searches for POMR 

Database/Source Keywords 

"uranium hexafluoride" OR "air count" AND "Piqua" accident AND "Piqua" AND Reactor "air dust" OR "air filter" OR "airborne test" AND "Piqua" AND Reactor "alpha particle" OR "belgian congo ore" OR bioassay OR bio-assay AND "Piqua" AND Reactor breath OR "breathing zone" OR BZ OR calibration OR "chest count" OR collimation OR columnation AND "Piqua" AND Reactor contamination OR curie OR denitration OR "denitration pot" AND "Piqua" AND Reactor derby OR regulus OR dose OR dosimeter AND "Piqua" AND Reactor dosimetric OR dosimetry OR electron OR environment AND "Piqua" AND Reactor "Ether-Water Project" OR exposure OR "exposure investigation" OR "radiation exposure" AND "Piqua" AND Reactor external OR "F machine" OR fecal OR "feed material" OR femptocurie OR film OR fission OR fluoroscopy AND "Piqua" AND Reactor "Formerly Utilized Sites Remedial Action Program" OR FUSRAP OR gamma-ray OR "gas proportional" OR "gaseous diffusion" AND "Piqua" AND Reactor health OR "health instrument" OR "health physics" OR "H.I." OR HI OR HP OR "highly enriched uranium" OR HEU AND "Piqua" AND Reactor hydrofluorination OR "in vitro" OR "in vivo" OR incident OR ingestion OR inhalation OR internal AND "Piqua" AND Reactor investigation OR isotope OR isotopic OR "isotopic enrichment" OR "JS 
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Table A2-2: Database Searches for POMR 

Database/Source Keywords 

Project" OR Landauer OR "liquid scintillation" AND "Piqua" AND Reactor log OR "log sheet" OR "log book" OR "low enriched uranium" OR LEU AND "Piqua" AND Reactor "maximum permissible concentration" OR MPC OR metallurgy OR microcurie OR millicurie AND "Piqua" AND Reactor "mixed fission product" OR MFP OR monitor OR "air monitoring" OR nanocurie OR "nasal wipe" OR neutron OR "nose wipe" AND "Piqua" AND Reactor nuclear OR Chicago-Nuclear OR "nuclear fuels" OR "nuclear track emulsion" OR "type A" AND "Piqua" AND Reactor NTA OR "occupational radiation exposure" OR occurrence OR "ore concentrate" OR "PC Project" AND "Piqua" AND Reactor permit OR "radiation work permit" OR "safe work permit" OR "special work permit" OR RWP OR SWP AND "Piqua" AND Reactor "phosphate research" OR photofluorography OR photon OR picocurie OR pitchblende OR "pocket ion chamber" OR PIC OR problem OR procedure AND "Piqua" AND Reactor radeco OR radiation OR radioactive OR radioactivity OR radiograph OR radiological AND "Piqua" AND Reactor "Radiological Survey Data Sheet" OR RSDS OR radionuclide OR raffinate OR reactor AND "Piqua" AND Reactor respiratory OR "retention schedules" OR roentgen AND "Piqua" AND Reactor sample OR "air sample" OR "dust sample" OR "general area air sample" AND "Piqua" AND Reactor 
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Table A2-2: Database Searches for POMR 

Database/Source Keywords 

"solvent extraction" OR source OR "sealed source" OR spectra OR spectrograph OR spectroscopy AND "Piqua" AND Reactor spectrum OR standard OR "operating standard" OR "processing standard" AND "Piqua" AND Reactor survey OR "building survey" OR "routine survey" OR "special survey" OR "technical basis" AND "Piqua" AND Reactor "thermal diffusion" OR "thermoluminescent dosimeter" OR TLD OR "Tiger Team" AND "Piqua" AND Reactor "tolerance dose" OR urinalysis OR urine OR "whole body count" OR WBC AND "Piqua" AND Reactor "working level" OR WL OR X-ray OR "X ray" OR Xray OR "x-ray screening" AND "Piqua" AND Reactor "Atomics International" AND "Piqua" DOE/CH-9111 OR DOE/CH-9225 

Table A2-3: OSTI Documents Ordered for POMR 

Document Number Document Title Requested Data 

N/A OSTI ID: 5179961 SRDB: 54696 

Piqua Nuclear Power Facility Radiological Surveillance Program 

10/29/2008 

N/A OSTI ID: 4666935 SRDB: 55945 

Piqua Nuclear Power Facility Monthly Operating Report No. 19 

10/29/2008 

N/A OSTI ID: 4614403 SRDB: 55944 

Piqua Nuclear Power Facility Monthly Operating Report No. 26 

10/29/2008 
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Table A2-3: OSTI Documents Ordered for POMR 

Document Number Document Title Requested Data 

N/A OSTI ID: 4592064 SRDB: 55943 

Piqua Nuclear Power Facility Monthly Operating Report No. 24 

10/29/2008 

N/A OSTI ID: 4181116 SRDB: 55942 

Piqua Nuclear Power Facility Monthly Operating Report No. 53 

10/29/2008 

NAA-SR-12561 OSTI ID: 4501434 SRDB: 55892 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 10, January 1--June 30, 1967. 

10/29/2008 

NAA-SR-12445 OSTI ID: 4799966 SRDB: 55883 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 9, July 1--December 31, 1966. 

10/29/2008 

NAA-SR-MEMO-12540 OSTI ID: 4496873 SRDB: 55938 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Report 55, June 1967. 

10/29/2008 

NAA-SR-4576 OSTI ID: 4143395 SRDB: 55904 

Piqua OMR Waste Gas Disposal 10/29/2008 

NAA-SR-8722 OSTI ID: 4123254 SRDB: 55926 

Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 2, Fiscal Year 1963 

10/29/2008 

NAA-SR-5688 OSTI ID: 4080480 SRDB: 55906 

Proceedings of the Organic Cooled Reactor Forum, October 

6-7, 1960 

10/29/2008 

NAA-SR-11628 OSTI ID: 4565167 SRDB: 55817 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 6, January 1-June 30, 1965 

10/29/2008 

NAA-SR-9473 OSTI ID: 4031136 SRDB: 55932 

Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 3, July 1, 1963-December 31, 1963 

10/29/2008 

NAA-SR-MEMO-10529 OSTI ID: 4639851 SRDB: 55934 

Preparation and Transportation of Piqua Fuel Elements. 

Summary Report 

10/29/2008 

NAA-SR-11142 OSTI ID: 4583353 SRDB: 55815 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 5, July 1-December 31, 1964 

10/29/2008 

NAA-SR-11650 Sect 3 Hallam Nuclear Power Facility and Piqua Nuclear Power 10/29/2008 
SEC-00126 04-27-09 Piqua Organic M

64 of 65 

Table A2-3: OSTI Documents Ordered for POMR 

Document Number Document Title Requested Data 

OSTI ID: 4582815 SRDB: 55818 

Facility 

NAA-SR-10307 OSTI ID: 4683534 SRDB: 55814 

Piqua Nuclear Power Facility Operations Analysis Program Progress Report No. 4, January 1-June 30, 1964 

10/29/2008 

NAA-SR-MEMO-12071 OSTI ID: 4468274 SRDB: 55935 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Report No. 41, January--April 1966 

10/29/2008 

NAA-SR-11995 OSTI ID: 4465693 SRDB: 55880 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 7, July 1, 1965--January 13, 1966 

10/29/2008 

NAA-SR-12148 OSTI ID: 4440848 SRDB: 55881 

Piqua Nuclear Power Facility Reactor Operations Analysis 

Program Semiannual Progress Report No. 8, January 1--June 30, 1966 

10/29/2008 

NAA-SR-11200 OSTI ID: 4571317 SRDB: 55816 

Aec Unclassified Programs. Quarterly Technical Progress Report, January-March 1965 

10/29/2008 

NAA-SR-3575 NAA-SR-MEMO-4048 OSTI ID: 4182273 SRDB: 55896 & 55940 

Preliminary Safeguards Report for the Piqua Organic Moderated Reactor (revised) Supplement I to the Preliminary Safeguards Report for the Piqua Organic Moderated Reactor (revised) 

10/29/2008 

NAA-SR-6893 OSTI ID: 4781060 SRDB: 55921 

Energy Absorption Rates in OMR Coolants 10/29/2008 

NAA-SR-MEMO-6581 OSTI ID: 4803723 SRDB: 55941 

Re-evaluation of Energy Absorption Rate in OMR`s 10/29/2008 

NAA-SR-12516 OSTI ID: 4799937 SRDB: 55890 

Safety Evaluation of PNPF Modifications 10/29/2008 

NAA-SR-3635 OSTI ID: 4221122 SRDB: 55901 

Preliminary Design Description for the Piqua Organic Moderated Reactor Plant, Piqua, Ohio (revised) 

10/29/2008 

N/A OSTI ID: 4343710 SRDB: 56036 

Recovery of Uranium from Piqua Cold Scrap Fuel Elements 

10/29/2008 

DPSPU-72-124-5 Safety Analysis Report: Piqua--Elk River Shipping Cask 10/29/2008 
SEC-00126 04-27-09 Piqua Organic M

65 of 65 

Table A2-3: OSTI Documents Ordered for POMR 

Document Number Document Title Requested Data 

OSTI ID: 4602683 SRDB: 55811 

(packaging of radioactive and fissile materials). Final Report. 

N/A OSTI ID: 4498453 

Piqua Nuclear Power Facility Surveillance and Recovery Program. Final Report 

10/29/2008 

N/A OSTI ID: 4683116 

Piqua Elk River Reactors Spent Fuel Shipping Cask Design Report 

10/29/2008 

N/A OSTI ID: 5446994 

Annual Survey of the Retired Piqua Nuclear Power Facility 10/29/2008 

DOE/CH/10370-1 OSTI ID: 6355044 

Annual Survey of the Retired Piqua Nuclear Power Facility 10/29/2008 

TID-18813 OSTI ID: 4702649 

Piqua-Elk River Reactors Spent Fuel Shipping Cask. Final Design Report 

10/29/2008 

NAA-SR-Memo-7676 OSTI ID: 4728186 

Piqua Operations Analysis Program. Annual Technical Progress Report, Fiscal Year 1962 

10/29/2008 

Table A2-4: Cincinnati Public Library Documents Ordered for POMR 

Document Number Document Title Requested Data 

N/A OSTI ID: 4815595 SRDB: 56595 

Piqua Nuclear Power Facility. 1. Status Report 2. General Description of the Piqua Reactor 

10/29/2008 

N/A OSTI ID: 4196389 SRDB: 56588 

Engineering Design of Piqua OMR 10/29/2008 

N/A OSTI ID: 4676966 SRDB: 53465 

Operating of Piqua Nuclear Power Facility from Criticality to First Scheduled Shutdown 

10/29/2008 

N/A OSTI ID: 4147910 

Safety Considerations for the Organic Moderated Reactor in Piqua, Ohio 

10/29/2008